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Surry Station Units 1 & 2, Final Safety Analysis Report
	ML19094A132
Person / Time
Site:	Surry
Issue date:	01/21/1970
From:	; Virginia Electric & Power Co (VEPCO)
To:	; US Atomic Energy Commission (AEC)
References
	Download: ML19094A132 (428)
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Vepco SURRY POWER STATION 1-*~- LI.NITS 1 AND 2 (f c'..,

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SAFETY ANALYSIS REPORT- :?.....-.-.._

   ,. VIRGINIA ELECTRIC AND POWER COMPANY
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i 12-1-69

TABLE OF CONTENTS OF THE FINAL SAFETY ANALYSIS .REP9RT 1 INTRODUCTION AND

SUMMARY

1.1 INTRODUCTION

1.2

SUMMARY

1.3 COMPARISON WITH OTHER STATIONS 1.4 COMPLIANCE WITH CRITERIA 1.5 COMMON AND SEPARATE FACILITIES 1.6 RESEARCH AND DEVELOPMENT 2 SITE j**- 2.1 2.2 GENERAL DESCRIPTION METEOROLOGY AND CLIMATALOGY 2.3 HYDROLOGY 2.4 GEOLOGY 2.5 SEISMOLOGY 3 REACTOR 3.1 GENERAL DESCRIPTION 3.2 DESIGN BASES 3.3 NUCLEAR DESIGN 3.4 THERMAL AND HYDRAULIC DESIGN AND EVALUATION 3.5 MECHANICAL DESIGN -~ 3.6 TESTS AND INSPECTIONS

ii 12-1-69 4 REACTOR COOLANT SYSTEM 4.1 DESIGN BASES 4.2 SYSTEM DESIGN AND OPERATION 4.3 SYSTEM DESIGN EVALUATION 4.4 TESTS AND INSPECTIONS 5 CONTAINMENT SYSTEM 5.1 GENERAL DESCRIPTION 5.2 CONTAINMENT ISOLATION 5.3 CONTAINMENT SYSTEMS 5.4 DESIGN EVALUATION. 5.5 TESTS AND INSPECTIONS 6 ENGINEERED SAFEGUARDS 6.1 GENERAL DESCRIPTION 6.2 'SAFETY INJECTION SYSTEM 6.3 *coNSEQUENCE-LIMITING SAFEGUARDS 7 INSTRU~NTATION AND CONTROL 7.1 GENERAL DESIGN CRITERIA 7.2 PROTECTIVE SYSTEMS 7.3 CONTROL SYSTEMS 7.4 NUCLEAR INSTRUMENTATION SYSTEMS DESIGN AND EVALUATION 7.5 ENGINEERED SAFEGUARDS INSTRUMENTATION 7.6 INCORE INSTRUMENTATION 7.7 OPERATING CONTROL STATIONS ,* 7.8 AUTOMATIC LOAD CONTROL . ?,. 9 COMPUTER

iii 12-1-69

8.1 8.2 8 ELECTRICAL SYSTEMS GENERAL DESCRIPTION AND

SUMMARY

DESIGN BASES 8.3 UTILITY SYSTEM INTERCONNECTIONS 8.4 STATION SERVICE SYSTEMS 8.5 EMERGENCY POWER SYSTEM 8.6 TESTS AND INSPECTIONS 9 AUXILIARY AND EMERGENCY SYSTEMS 9.1 CHEMICAL AND VOLUME CONTROL SYSTEM 9.2 BORON RECOVERY SYSTEM 9.3 RESIDUAL HEAT REMOVAL SYSTEM 9.4 COMPONENT COOLING SYSTEM 9.5 FUEL PIT COOLING SYSTEM 9.6 SAMPLING SYSTEM 9.7 VENT AND DRAIN SYSTEM 9.8 COMPRESSED AIR SYSTEMS 9.9 SERVICE WATER SYSTEM 9.10 FIRE PROTECTION SYSTEM 9.11 WATER SUPPLY AND TREATMENT SYSTEMS 9.12 FUEL HANDLING SYSTEM 9.13 AUXILIARY VENTILATION SYSTEM 9.14 DECONTAMINATION FACILITY

iv 12-1-69 10 STEAM AND POWER CONVERSION 10.1 GENERAL DESCRIPTION 10.2 DESIGN BASES 10.3 SYSTEM DESIGN AND OPERATION 11 RADIOACTIVE WASTES AND RADIATION PROTECTION 11.1 GENERAL DESCRIPTION 11.2 RADIOACTIVE WASTE SYSTEMS

11. 3 RADIATION PROTECTION 12 CONDUCT OF OPERATIONS 12.1 GENERAL 12.2 ORGANIZATION 12.3 TRAINING 12.4 SHIFT PERSONNEL 12.5 HEALTH PHYSICS 12.6 OPERATIONS PROCEDURES 12.7 RECORDS 12.8 REVIEW AND AUDIT OF OPERATIONS 12.9 INSERVICE INSPECTION 13 INITIAL TESTS AND OPERATION 13.1 TESTS PRIOR TO INITIAL REACTOR FUELING 13.2 FINAL STATION PREPARATION 13.3 INITIAL TESTING IN THE OPERATING REACTOR.

13.4 OPERATING RESTRICTIONS

V 12-1-69

- 14.1 GENERAL 14 SAFETY ANALYSIS 14.2 CORE AND COOLANT BOUNDARY PROTECTION ANALYSIS 14.3 STANDBY SAFEGUARDS ANALYSIS 14.4 GENERAL STATION ACCIDENT ANALYSIS 14.5 LOSS-OF-COOLANT ACCIDENT 15 STRUCTURES AND CONSTRUCTION 15.1 STRUCTURES AND MACHINERY ARRANGEMENT 15.2 STRUCTURAL DESIGN CRITERIA 15.3 MATERIAL

., 15.4 15.5 15.6 CONSTRUCTION PROCEDURES AND PRACTICES SPECIFIC STRUCTURAL DESIGNS OTHER CLASS I STRUCTURES APPENDIX A - REPORT, SITE ENVIRONMENTAL STUDIES, SURRY POWER STATION APPENDIX B - SEISMIC DESIGN FOR THE NUCLEAR STEAM SUPPLY SYSTEM TECHNICAL SPECIFICATIONS 1.0 DEFINITIONS 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 3.0 LIMITING CONDITIONS FOR OPERATION 4.0 SURVEILLANCE REQUIREMENTS 5.0 DESIGN FEATURES 6.0 ADMINISTRATIVE CONTROLS

8-i 12-1-69 TABLE OF CONTENTS Section Title

                        -  -                     Page 8       ELECTRICAL SYSTEMS                     8.1-1 8.1       GENERAL DESCRIPTION AND

SUMMARY

8.1-1 8.2 DESIGN BASES 8.2-1 8.3 UTILITY SYSTEM INTERCONNECTIONS 8.3-1 8.4 STATION SERVICE SYSTEMS 8.4.1-1 8.4.1 4,160 VOLT SYSTEM 8.4.1-1 8.4.2 480 VOLT SYSTEM 8.4.2-1 8.4.3 120 VOLT VITAL BUS SYSTEM 8.4.3-1 8.4.4 125 VOLT D-C SYSTEM 8.4.4-1 8.4.5 LIGHTING SYSTEM 8.4.5-1

8.5 8.6 EMERGENCY POWER SYSTEM TESTS AND INSPECTIONS 8.5-1 8.6-1

8. 1-1 12-1-69

8.1

8. ELECTRICAL SYSTEMS GENERAL DESCRIPTION AND

SUMMARY

The electrical systems include the equipment and the systems necessary to gen-erate power and deliver it to the high voltage switchyard. It also includes facilities for providing power and controlling the operation of electrically driven auxiliary equipment and instrumentation. The main electrical connec-tions are shown in Figure 8.1-1. This section describes the electrical system for Unit No. 1. The Unit No. 2 ., electrical system is identical but completely independent of Unit No. 1. reserve station service transformer banks are common to both Unit No. 1 and Unit No. 2, and are sized to start*upa single unit or shut downboth units. The The output of the main generator is fed into and operated as an integral part of the Virginia Electric and Power Company system. The operation of the automatic load control is discussed in Section 7.8. The normal station service power system is designed to provide continuous power to the station auxiliaries during periods of generation and to transfer auto-matically to the reserve station service to ensure continued power to equipment when the main generator is off the line. Critical instrumentation is fed from a reliable and stable vital bus system to

ensure continuous monitoring and control of critical instrument channels.

8.1-2 12-1-69 In general, all auxiliaries of major equipment will be connected to the same power distribution branch as the major piece of equipment, the only exceptions being those appurtenances that will. be required to operate when the major piece of equipment shuts down. An example is a pump discharge motor operated valve that must close when the ptllllp stops. Station batteries are provided for circuit breaker control power, emergency lighting and operating power for vital equipment until normal power is restored or onsite emergency generation is available.

                                                                                                                                                           -I 0

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                                                               <                    HYDRO                            BRG COOL PP (UNIT-,1 ONLY)                               ::0 ITI (J)    l> ITI        TEST PP      CD-v                WATER PP               0

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8. 2-1 12-1-69 I 8.2 DESIGN BASES The electrical systems are designed to supply electrical power to all essential unit equipment during normal station operation and under incident conditions.

The electrical system components vital to unit safety, including the emergency diesel generators, are designed and protected as necessary so that their integrity is not impaired by potential earthquakes, high winds, floods or dis-turbances on the external electrical system. Cables, motors and other elec-trical equipment required for operation of the Engineered Safeguards are suitably protected against the effects of either a nuclear system accident or a* severe external environmental phenomenon, in order to ensure a high degree of reliability. The main generator, described in Section 10.3.3, feeds electrical power at 22 kv through an isolated phase bus to the main step-up single phase trans-formers and the unit station service transformers located adjacent to the turbine building. The primary side of each station servi*ce transformer, three per unit, is connected to the unit isolated phase bus at a point between the generator terminals and the low voltage connection of the main step-up trans-formers. During normal operation, station auxiliary power is taken from these transfonners which are rated 22 kv-4.16 kv. They supply three independent 4,160 v auxiliary buses and are designed to limit the short circuit fault duty on any one bus to within the interrupting capability of the 250 Mva air circuit breakers. The primary sides of the reserve station service transformers are connected to the 34.5 kv tertiary winding of either of two autotransformers in the high

8. 2-2'.

12-1-69 voltage switchyard as described in Section 8.3. The reserve station service transformers and their connection with the autotransformers are the "standby" power sources mentioned in Section 1. During startup, shutdown, or hot standby conditions, station auxiliary power is taken from these reserve station service transformers which have 4.16 kv secondary windings so as to provide an alternate supply to the three 4,160 v auxiliary buses. The screen well area is supplied through either of two 34.5 kv to 4.16 kv transformers, each of which is supplied through underground cable by the tertiaries of the autotransformers in the switchyard. Essential electrical equipment components are specified to withstand, without loss of function, the maximum conditions expected during normal operating and post-incident environments, and during operatic,n of the safeguard equipment during the incident. It is expected that the maximum incident conditions within the containment will be 280° Fat 40 psia for 30 min. Test data supplied by the equipment vendor confirms that the component or system can survive the post-incident environment. Should suitable equipment not be available, the detail plant design incorporates features to modify the environment to be compatible with the equipment. In the containment, essential electrical components and conductors are pro-tected from the forces generated during an incident by group separation. By physically separating each group and providing conductor barriers where neces-sary, the failure of one group does not jeopardize any other group. In the case of multiple instrument channels in one location, such as the channels

8.2-3 12-1-69

associ~ted with the single pressurizer, physical separation is carried out as far as possible and t_he circuitry arranged so that multiple instrument failures are always in the safe direction. Electrical cable connections are run from the instrument transmitter to the area outside the crane support wall using the shortest path while providing separation between redundant channels. The crane wall acts as a further barrier against any forces *generated during an incident.

The 480 v system is divided into three double ended bus sections, and each sec-tion is fed from a counterpart 4,160 v bus through individual 4,160 v-480 v station service transformers. In general, the 4,160 v and 480 v switchgear are of metal-clad dead front con-struction with closing and tripping control power taken from the station batteries

Each starter or breaker cubicle is isolated from the adjacent cubicle with metal
barriers and each bus section is physically separated from all others.

All switchgear associated with Engineered SafegOtlrds equipment is separated from the main switchgear area and is readily accessible in the_ main control area. Power and control* cables are distributed from the s:witchgear and control area by means of rigid metal conduits or ladder.type cable trays. All connections at the 22 kv voltage level are made with isolated phase con-struction designed for forced air cooling. Fire resisting fillers, tapes, binders and jackets were specified for all cable

\. construction.

8.2-4 5-31-71 All cable tra~ installations have approved fire stops in both horizontal and vertical runs and are provided with solid raised or a corrogated solid aluminum cover. Covers may be omitted on top trays run under solid floors. All conduit installations consist of plastic conduit encased in concrete or exposed rigid metal conduit. The main feeds to the 4,160 v switchgear from the unit station service trans-formers are shielded single conductors, with neoprene jackets installed in ladder type trays with 1. 25 diameter spacings between conductors. The main feed to the 4,160 v switchgear from the reserve transformer are of the same construc-tion as the unit transformer feeds except that part of the run length will be in a duck bank. One reserve transformer feeder has separate routing to the 4,160 v switchgear, physically isolated from all other transformer secondary leads. For all leads supplying Engineered Safety equipment, the cable is 3/ c with interlocked armor overall, run in ladder trays or properly mounted and supported when run external to ladder trays, or 3-1/c triplexed, run in conduit, with the exceptions of the 480 volt equipment supplied from motor control centers and the emergency generator leads. The only 480 volt exceptions are for 30 hp motors and smaller. The emergency generator leads entering the emergency switchgear room from the duct bank have been derated for cable in conduit in accordance with IPCEA standards. In the emergency switchgear room, some of the cables have been run in trays. These ladder type trays have solid covers placed directly on the top of the trays and with a* solid transite or asbestos blanket placed on the bottom of the

8.2-5 10...:15-70 trays prior to installation of cables. This installation has the same

protection integrity as cable in conduit and facilitates installation and inspection of these critical cables.

Control cables are of single or multi-conductor construction with cross-linked polyethylene insulation rated at 1,000 v and with overall flame retardant jacket. Low voltage instrument connections are made using flame retardant insulated cables , rated at 600 v. These* cables are provided with a tot al cover age electrostatic shield and an overall flame retardant jacket. The normal current rating of all insulated conductors is limited to that con-tinuous value which does not cause excessive insulation deterioration from heating. Selection of conductor s.izes are based on "Power Cable Ampacities," published by the Insulated Power Cable Engineers Association (IPCEA).

Four vital buses are provided for critical instrumentation and reactor protec-tion circuits. Two vital buses are each fed from an independent station battery through static inverter and are maintained at 120 v, +/-6 v, and 60 cycles,

 +/-0,5 cycles. The remaining two vital buses are fed through independent sola regulating transformers and energized from separate 480 v emergency buses.

The vital bus system is described in Section 8.4.3. The enclosures for motors and electrical switchgear are selected to suit the local conditions and are designed in accordance with specifications issued by the National Electrical Manufacturers Association (NEMA)

8.2-6 12-1-69 The station batteries are sized to operate circuit breaker controls, turbine shutdown oil pumps, instrumentation, emergency lighting and vital nuclear channels for 2 hr without benefit of any station power. The battery chargers are connected to the emergency bus and provide charging current to the battery and load when the emergency bus is energized.

Lighting distribution and intensities have been selected in accordance with the latest recommendations of the Illumination Engineering Society (IES). All electrical equipment and cables are designed to operate within their normal rating or temperature rise. The servic.e factor or overload rating will not be used during normal operating conditions.

8.3-1 2-15-71

8.3 SYSTEM INTERCONNE,~Jf~NS

                                                                                      ,l, Unit 1 and Unit 2 are connected with the Virginia Electric and Power Company
   ,system at a transmission substation near the station. The connection is essentially a double one since it is made through both the 230 kv and 500 kv transmission systems.

Each main generator feeds electrical power through an isolated phase bus to a bank of three single phase transformers provided for each unit, thereby stepping the generator voltage up to transmission voltage. Electrical energy generated by Unit 1 at 22 kv is raised to 230 kv by the main

transformer and delivered to the 230 kv switchyard. Electrical energy generated by Unit 2 at 22 kv is raised to 500 kv' by the main transformer and delivered to.

the 500 kv switchyard. Figure 8.3-1 is a single line diagram of the transmission substation for Surry Power Station. Figure 8.3~3 is a simplified single iine diagram*which shows the connections betw~en the transmission substation, Figure 8.3-1, and the power station one line diagram, Figure 8.1-l. I Station service transformers connected to the isolated phase bus from each main generator normally supply power to the auxiliaries of each unit at 4,160 v. During start-up and emergencies, reserve station service power for the auxiliaries of either unit is supplied from tertiary windings of two 500/230 kv transmission intertie autotransformers which connect the 500 kv and 230 kv sections of the substation

8.3-2 12-1-69 The 230 kv sw:Ltchyard is of the "breaker and a half" design with facilities for six 230 kv lines in service with Unit 1 and seven 230 kv lines in service with Unit 2. The 500 kv switchyard is also of the breaker and a half" layout.

Initially, wh,an Unit 1 is placed in service, it will be connected to a single 500 kv line by one of the autotransformers. ,When Unit 2 is placed in service, the SOO 1tv substation will be expanded to a five position ring bus with the connections being the Unit 2 gen~rator, two 500 kv lines, and two 500/230 kv autotransformers. With Unit 1, both S00/230 kv autotransformers are in service to supply reserve station power from the 34.5 kv tertiary windings. Initially one autotransformer is connected to the 500 kv and 230 kv systems and the other connected at 230 kv only, but, with the addition of Unit 2 the connection is completed to the 500 kv substation. The 500 kv and 230 kv systems are generally independent and provide alternate sou.rces of reserve station power. The substation can be expanded for future units and lines as required. Two gas turbi.nea are installed at the Surry site East of the 230 kv substation. One unit is rated at 16 Mw and the other at 25 Mw * .These units are a part of the Virginia Elec.tric and Power Company system and are primarily used for load peaking. One.of ~e units has a black start capability with a start up time of approximately 10 minutes. The units can be operat~d locally or from a remote panel located in the Main Control Room at the Surry Power Station. Supervisory equipment provides alarm, control, and indication so that the two gas turbines can be started, loaded, and

trirped from the Main Control Room *.

FIG.8.3-3 DEC. I 1969 TRAN SM ISSI ON LINES TO 230 KV SYSTEM FOR SWJTCHYARD DETAIL SEE FIG. 8.3 TRANSMISSION LINES TO 500 KV SYSTEM

                                                                                                                                  *I 13.8 KV h                 I I r 1, :

L( I I I

                                                                                                                            .... _J    r GAS TURBINES                                                                                              L.J r~

I r,

                                                                                                                                         +.:-~- -------,

L,J L,.J I I 500 KV BUS TO SCREENWELL TO SCREENWELL TRANSFORMER r----t*TRANSFORMER 34.5 KV 34. 5 KV SWJTCHYARD BUS NO. 2 SWI.TCHYARD BUS NO. I AREA N.O. AREA

                +                                                                                                                                                                         l I                                                                                                                                                                         I

--~ L ____ _ I I y POWER STATION AREA MA IN TRANSFORMER NO. I (3-J,6)

                                                                                                                                               ---------~------~I MA IN TRANSFORMER '\/\YV' N 0. 2 ( 3 - I II° l

I
I I

I I f POWER STATION AREA STA. RES. ST1~. RES. STA. RES. SERV. STA. SERV. STA. SERV. STA. TRANS. TRANS. TRANS. TRANS. TRANS. TRANS. I I I 4160 V 4160 V 4160 V I BUS BUS BUS I 3 ,..+'\ ( __.l l

\'

N.O. N.O. N.O. ..... _,. ) MAIN MAIN GENERATOR GENERATOR NO. I TO UNIT NO. 2 NO. 2 TO 4160 V EMERGENCY FOR DETAJ LS OF 4160 V SYSTEMS TO 4160 V EMERGENCY BUS NO I SEE FIG. 8.1 BUS NO. 2

- ELECTRICAL INTERCONNECTIONS BETWEEN* THE SWITCHYARD AND POWER STATLON SURRY POWER STATION

./ -

                                             ---------~*//
                                                                    /
                                                                        /
                                                                           //*r...-t I

II I \

SF FIG_ 8.3-2 DEC. I, 1969 MONOHGAHEt.A/ \ I \ KJ10I\AC El£Ul.'.C I \ POwER COW>ANr I J POWER I ICOMPANY-- I LEGEND J = - : - TFtAMSMISSION LINES, !100 KV I I - TRANSMISSION LINES, 23~1:!&-11&-159 KV I - - DISTRIBUTION LINES, 46TO 23 KY I - - Dl5TR1BUT10N LINES BELOW.Z3KV I * * * *

OTHER UTILITY SVSTEMS I ......... PRINCIPAL INTERCONNECTIONS I
STEA.Iii GENERATING STATION J 0 HVORO GENERATING STATION I t:,. PRINCIPAL SUBSTATIONS I Q INCORPORATED TOWNS AMO CITIES SERVED I Efl COMMUNITIES SERVED AT WHOLESALE

                                                        ',                                                            II     U.S. GOVERNMENT HYDRO
                                                              ',,

STEAM GENERATING STATION (NUCLEAR)

_ /_.-*-**-*+ 0 ,. 30 40 eo

          ~==:0:-.:0.-.-.-.-:.-:.~-----*.
 ,,ila--,

19-- TRANSMISSION LINE MAP SURRY POWER STATION

8.3-3 12-1-69

The two generators are connected in parallel to the low voltage side of a 13.8/230 kv transformer. Each generator has a breaker which is .used for synchronizing and tripping. The high voltage side of the transformer is connected to No. 2 230 kv bus.

Transmission system connections for Unit l consist of the following lines which are an integral part of the Vepco transmission system:

1. One 500 kv line to Elmont substation near Richmond, Virginia.

2. Two 230 kv lines to Hopewell substation near Hopewell, Virginia.

3. One 230 kv line to Suffolk substation near Suffolk, Virginia.

It will connect to two 230 kv lines going to North Carolina and

4.

5. one 230 kv line to the Norfolk area. One 230 kv line to Churchland substation in Portsmouth, Virginia. One 230 kv line to Newport News substation in Newport News, Virginia.

6. One 230 kv line to Whealton substation in Hampton, Virginia.

Additional transmission system connections for Unit 2 consist of:

 .1. One 500 kv line to Carson substation near Petersburg, Virginia.

2. One 230 kv line to Greenwich substation in Virginia Beach, Virginia.

The transmission lines leave the high voltage substation along two main rights of way. Each right of way includes transmission lines which principally route toward East and West locations in the Virginia Electric and Power Company system

SF 8.3-4 12-i-69 The transmission system can handle the full output of both units at Surry upon the loss of any two transmission circuits connected to the Surry substation.

Figure 8.3-2 is a location map showing Surry Power Station, the associated transmission lines, and their system connections.

8.4.1-1 12-1-69

- 8.4 8.4.1 STATION SERVICE SYSTEMS 4,160 VOLT SYSTEM Alternating current station service power is distributed from the 4,160 v switchgear. This switchgear is energized from the main generator and unit station service transformers ~uring normal operation, or from the reserve station service transformer source during startup, hot standby, or shutdown operation {Section 8.2).

The 4,160 v system is duplicated for Unit No. 2 The 4,160 v switchgear is arranged in three independent bus sections. Each bus section has a capacity of about 3,000 amp. Each feeder or motor circuit is protected by overcurrent relays which trip the associated breaker for a sustained overload or fault. During unit startup, the total alternating current demand of the unit is supplied from the reserve station service source. After the unit has attained operating. conditions and the turbine generator synchronized and connected to the system, the station service load is transferred to the unit statiop service.tran~formers. This transfer is performed without a power interruption by momentarily feeding the 4,160 v switchgear from both the reserve and unit transformers. The reserve station service transformer is then disconnected and the turbine generator will supply its own auxiliaries. Loss of normal supply to any bus section automatically trips the normal source

breaker and closes the alternate source breaker. Unit operation with two loops is possible with one section of 4,160 v switchgear out of service.

8~4.1-2 10-15-70 Motors larger than approximately 300 hp are operated at 4,160 v and are arranged for across the line starting. One circuit from each bus section feeds two 4,160/480 v station service transformers. The 480 v system is described in Section 8.4.2. Two independent sections of emergency 4,160 v bus and switchgear are provided. Each section is sized to carry 100 per cent of the emergency load. These emergency sections are energized from the reserve station service transformer during normal operation, startup and shutdown. In the eventlof total loss of station auxiliary power, the emergency 4,160 v buses are isolated from the normal supply and energized from the emergency generators as described in Section 8.5. A manually operated air circuit breaker position is provided so that a 4,160 v emergency bus section may be connected to the redundant emergency bus section. This feature will be used for maintenance or abnormal conditions only, and will be under administrative control. This breaker will be removed from the cubicle and will not be installed when the unit is operating.

8. 4. 2-1 12-1-69 8.4.2 480 VOLT SYSTEM The 480 v station service system distributes and controls power for all 480 v and 120 v a-c station service demands. The source of power for the 480 v system buses is from the counterpart 4,160 v system buses. This system is shown in Figure 8.1-1.

The switchgear is metal-clad, with 125 v d-c operated air circuit breakers and arranged with six independent bus sections. The 4,160/480 v transformers are air cooled. The transformers are throat connected to the switchgear. Normal operation is with the bus sections independent of each other. Motors up to approximately 300 hp are connected to the 480 v switchgear. Reduced unit out-put is possible with two 480 v bus sections out of service. Motor control centers are located throughout the station and are used for 480 v power distribution and control of motors smaller than 100 hp. The motor control centers are fed from 480 v switchgear buses through circuit breakers. Engineered safeguards equipment items operating at 480 v are fed from independent 480 v buses and switchgear that are energized from either the reserve station service power or the emergency generators.

8.4.3-1 12-1-69 8.4.3 120 VOLT A-C VITAL BUS SYSTEM There are four separate 120 v a-c vital buses, two of which are supplied by independent 10 kva, single phase, static inverters; and two by independent 10-kv, sola regulating transformers as shown in Figure 8.4.3-1. Normal load on one vital bus cabinet is approximately 5 kva. The two inverters are each connected to one of the station batteries. Normally the inverter load is absorbed by the battery charger. Upon charger.failure, the battery will pick up the inverter load. Manual switches are provided to permit one inverter to feed two distribu-tion cabinets.

The inverter outputs are regulated automatically at 120 v a-~

  +/-6 v and 60 Hz+/- .5 Hz. The sola regulating transformers are fed from ind~pendent 480 v emergency buses and each transformer feeds one cabinet *

- The vital buses constitute a very reliable electrical system, and with four supplies, provides a stable source of power to vital instruments and equipment and thus eliminates spurious shutdowns and guarantees proper action when power is required.

The normal power source to the vital bus inverter comes from the station alter-nating current supply. Should the normal power source fail completely orbe subject to transient voltage or frequency variations, the vital bus inverter automatically transfers to the station battery. This static transfer takes place without disturbing the vital bus voltage or frequency. --~-*-** The output.of each inverter is connected to .a distribution cabinet through a normally closed air circuit breaker. The distribution cabinets have 15 and 20 amp

8.4.3-2 2-15-71 branch circuit breakers to feed reactor protection and other vital instrument channels. Reactor protective scheltee have redundant channels and the power sources are provided from redundant vital bus cabinets. The 120 v, 60 Hz output from each inverter and the sola regulating transformers is ungrounded. Instrumentation is provided to detect an accidental ground. A ground on one phase does not interrupt service and, with multiple channels, it is possible to correct the ground fault without tripping the reactor or sacri-ficing protection. Because-of the fail safe circuitry of the reactor protective instrumentation, a power source failure to an instrument channel results in a reactor trip signal from the affected channel. Multiple power supplies are provided to prevent a common power supply failure from initiating a false reactor trip. The vital bus rectifiers and inverters are assembled from high quality components, conservath,ely designed for long life and continuous operation. By avoiding the use of electromechanical devices, routine maintenance downtime is greatly reduced. There are no vacuum tubes or moving parts in the completely static vital bus supply systems. Magnetic amplifiers, transistors, and silicon rectifiers are used to provide trouble .free operation.

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8.4.4 125 VOLT D-C .SYSTEM The 125 v d-c batteries supply power for operation of turbine generator emergency auxiliaries, switchgear, annunciators, vital bus inverters and emergency lighting, as shown in Figure 8.4.4-1. The principal equipment items in this system are two 60-cell lead-acid batteries, two static battery chargers, and two battery distribution switchboard.s. A separate battery, battery charger, and distribution switchboard will be available for use in the screenwell structure.

The batteries are of the central power stat:lon type and are designed for continuous duty. Each battery consists of 60 cells connected in series . Each cell is of the sealed type, assembled in a shock absorbing, clear plastic container, with covers bonded in place to form a leakproof seal. The batteries are mounted on protected, corrosion resistant, earthquake resistant racks for security and to facilitate maintenance. The two battery areas are separated from each other and from the switchgear room. Normally, the two battery bus sections are operated independently, with the bus tie breaker open. Each charger supplies power for operation of equip-ment connected to that bus section and maintains a floating charge on its associated battery. A manually operated bus tie breaker provides for parallel operation of the chargers and batteries or operation with either battery or charger out of service for maintenance.

8.4.4-2 5-1-71 The four static battery chargers (2 per 125 v d-c bus) are identical, each having an output of 200 amp at 132 v d-c with an input of 440 v a-c, 3 phase. Each charger is equipped with a direct current voltmeter, ammeter, ground detector, alternating current failure relay, and low charging current alarm relay. Loss of a-c or low charging current is alarmed in the Main Control Room. Battery ground indicators are located in the Main Control Room. Battery voltage is indicated to the operator on the main control board and continuously recorded on recorders located in the Control Room. The battery chargers are energized from emergency motor control centers. The battery distribution switchboards are NEMA Class II metalclad structures, each with a 125 v d-c, 2-wire underground main bus, and 2-pole manually operated air circuit breakers. During normal operation, the 125 v d-c load is fed from the battery chargers with the batteries floating on the systems. Upon loss of station alternating current power, the entire direct current load is drawn from the batteries. The power load imposed on each battery will be initially high. After the turbine generator has coasted to a stop, the direct current auxiliary bearing and seal oil pump motors may be stopped. This removes a motor load of approximately 75 hp from the batteries. The remaining load consists primarily of emergency lighting and vital bus inverters. The batteries are sized for 2 hr operation after which it is assumed that station power or emergency generation power will be available to energize the battery chargers.

8.4.5-1 12-1-69

8.4.5 LIGHTING SYSTEM Normal lighting for turbine areas, reactor containments, auxiliary building, fuel building, and service buildings are provided from local lighting cabinets located in the area of service. These cabinets are fed from a double ended lighting switchboard which is energized from 2 independent 250 kva, single phase, 4,160 - 240/120 v dry type, self ventilated transformers. Normally the two buses of the double ended switchboard are separate. They are capable of being tied together if one transformer fails.

Normal lighting for the office building and remote areas is supplied through local 480 - 120/240 v, single phase, drytype transformers. Emergency lighting

for remote areas is provided by local self-contained battery-powered emergency lighting units.

Emergency lighting for turbine areas, auxiliary building, fuel building and service building is in the form of incandescent units and are normally de-energized. These lights are automatically switched to the d-c system upon sensing loss of voltage on the lighting switchboard. Emergency lighting for the reactor containment is energized at all times from an independent d-c circuit. The turbine room operating floor also has an independent feed from the battery and automatically comes on if lighting intensity falls below certain levels. Emergency lighting feeds are provided. from both units to the Main Control Room

and turbine room mezzanine to provide the best possible protection.

8.5-1 12-1-69

8.5 EMERGENCY POWER SYSTEM The Emergency Power System is an onsite, independent, automatically starting power source. It supplies power to vital auxiliaries if a normal power source is not available.

The Emergency Power System consists of three diesel generators for the two units. One generator is used exclusively for Unit 1, the second for Unit 2 and the third unit functions as a *back-up for ei.ther Unit l or 2. Each engine generator has 100 percent capacity and is connected to independent electrical buses. Each unit has two emergency buses normally fed from independent reserve 4,160 v systems, with the diesel generators functioning as an onsite back-up

system.

The diesel eng'ine starting circuitry accepts the following signals:

1. Undervoltage on emergency bus

2. Safety injection signal

3. High Containment pressure signal

4. Manual All safeguards equipment items are duplicated and connected to separate emer-gency buses. In the event of an equipment failure on one emergency bus coin-cident with an emergency generator failure on the other bus, it is possible to connect both electrical buses to one generator. This emergency connection

is made by manual operation of the bus tie breaker.

SF 8.5-2 12-1-69 Each diesel generator is reliable in operation. by u,se of redundant components as follows: This reliability is achieved *

1. Fuel System Duplicate fuel systems with independent fuel and transfer pumps, strainers and filters are provided.

2. Air Supply Starting System Duplicate air starting systems with independent compressor, valves and accumulators are provided~

3. Control Storage Battery 4.

Each unit has its own independent control storage battery, Control Equipment Each unit has individual control panels, metering, regulation and excitation equipment~ The diesel generators and associated equipment are located in a Class I and tornado protected structure. Each generator and its associated equipment will withstand, without loss of function, either the Design Basis Earthquake or the atmospheric pressure drop associated with the design tornado. Each emergency diesel generator set is sized to start and accept the post-design basis accident load in about 10 sec, be loaded with the requisite loads within about 25 sec, and carry the maximum emergency load envisioned for each unit. The starting, accelerating and loading times of the emergency

SF 8.5-3 12-1-69

generators using simulated loads have been witnessed and checked before the units were accepted from the engine manufacturer.

accept load is te*sted as described in Section 8.6. The continued ability to Each emergency bus provides power to the following safeguards equipment. All items are not required simultaneously but are automatically or manually pro-grannned to start as needed.* 1 - Containment spray pump 1 - High head safety injection pump 1 - Low head safety injection 1 - Inside recirculation spray pump

1 - Outside recirculation spray pump 1 Containment vacuum pump 1 - Motor control center for valves, air compressor, fuel oil transfer pump, etc.

1 - Emergency steam generator feedwater pump The failure of any item to operate is detected and the corresponding item on the alternate bus is started. Each emergency generator is provided with a starting system sized for three engine starts without outside power. Each engine also has an independent fuel tank with capacity for approximately 3 hr of full load operation. These inde-pendent fuel tanks are filled by transferring fuel from a buried and tornado

8.5-4 12-1-69 protected main fuel oil storage tank. Two motor driven fuel oil transfer pumps, one of which is _standby, are powered from the emergency generator buses to ensure that an operating emergency generator has a continuous supply of fuel. If any saf,:guard equipment fails to operate ~iutomatically, manual operation is possible, remotely from the Main Control Room or locally at the switchgear. The switchgear for each emergency generator is physically and electrically isolated from each other. If the loss of normal power is not accompanied by a loss-of-coolent incident, the safeguards equipment is not required. Under this condition, other plant auxiliary equipment, such as a component cooling pump, residual heat removal pump, etc., may be operated manually up to the capacity of the emergency gen-erators. Instrumentation is provided to indicate generator loading. Each piece of vital equipment is connected to the auxiliary electrical power system with an exclusive circuit. Each circuit has an air circuit breaker with overcurrent protection, and a control switch with red, amber, and green indicating lights mounted in the Main Control Room. The red light shows that the power circuit is energized, trip coil continuity is present, and control power is available. The green light is lit when the power circuit is de-energized and monitors the availability of control power. Simultaneous lighting of the amber and green lights indicates an automatic trip of a feeder or source circuit. Major items have ammeters to indicate circuit current.

8.5-5 12-1-69

Isolation of a failed circuit is automatic and is identified by the indicating lights in the Main Control Room. All automatic tripping functions also energize an audible signal to alert the control room operator. Individual pro-tective relays have signal targets to indicate that automatic operation has taken place.

The voltage level and current loading of all station distribution buses are dis-played in the Main Control Room. The status of the switchyard breakers and the source of reserve station power are readily available to the operator. Indi-cating lights show the source of power to each bus. Alternate sources may be manually selected by the operator but prearranged automatic transfer takes place on failure of the normal source *

The emergency diesel generator panel contains instruments and controls to serve the emergency bus. Provisions for synchronizing the emergency generator man-ually with the normal station service power systems are also provided. The emergency generator is manually synchronized with the system and loaded for periodic load tests.

Emergency switchgear is located in the shielded control area below the Main Con-trol Room. The status of the emergency power bus can be determined at the emer-gency switchgear. Emergency distribution air circuit breakers can be manually operated at the switchgear. (Reference Section 7.7) All routine control of normal and standby electrical power is from the Main Con-

trol Room. However, essential loads can also be controlled from the emergency

8,5-6 12-1-69 switchgear located within the shielded confines below the Main Control aoom~ The emergency switchgear is designed so that local operation is possible with o-r. without control power. The electrical power distribution system for the Surry Power Station provides duplicate systems for emergency components. Each system is continuously ener-gized from the external system grid or from onsite emergency generators. The automatic transfer from normal to alternate source, when required, is accom-plished by opening the normal source air circuit breaker and closing the emer-gency generator air circuit breaker. The electrical power distribution system is normally supplied from the station service transformer which is connected to the unit main generator. Automatic bus transfer occurs when the undervoltage of the station service bus is detected.

The emergency bus and the supply for all essential components are normally con-nected to the reserve station supply. The standby generator is started on the receipt of an undervoltage signal on the emergency bus, safety injection signal or containment high pressure signal. A loss of reserve station supply to the emergency generator bus opens the normal circuit breaker and closes the standby generator breaker. At the same time an independent circuit trips or sheds the two nonessential pumps connected to each emergency bus. If for some reason these pump breakers do not trip, another circuit opens the bus tie breaker that connects a short bus section to the emergency bus.

The emergency generator and the normal station service are never synchronized, except during periodic test. synchronizing is not used. This synchronizing is done manually; automatic

 ~--- ---

8.5-7 10-15-70 All essential electrical components and circuits are located and distributed within protected zones. All cables, conductors, motors, pumps, control stations, etc., are identified by a mark number or by function. The markings consist of painted stencils or marked tags applied or attached to each component.

All control switches on the main control board are clearly identified by system. Emergency switchgear and control centers are identified as control devices for essential components. All switchgear associated with the electrical feeds to the emergency buses are enclosed in metal housing and protected from the weather. These enclosures are equipped with thermostatically controlled electric heaters to prevent condensation. The majority of lines and valves required for containment isolation or engineered safeguards operation are located in protected areas. All lines or valves subject to freezing or ,cry~-1:~~-l~_:rntion of boron are _electrically~ heat traced and insulated. The heat tracing is automatically energized when the temperature drops below 35°F for water lines or 170°F for boron lines. Therefore, icing or crystailization could ndt

interfere with the isolation of containment or the i,nj ec tion of coolant during inc id en t conditions .

8.6-1 -~ 12,-1-69 8.6 TESTS AND INSPECTION. All electrical equipment was specified for manufacture in ~trict accordance with the latest requirements of NEMA (National Electrical Manufacturers Association), IEEE (Institute of Electrical & Electronic Engineers), or ANSI (American National Standards Institute, Inc.) standards, where applicable. Electrical equipment was protected during shipment and was properly stored at the job site during construction. Installation of all equipment was under the supervision of a qua.;tified elec-trical construction engineer. Special attention was given to,mechanic~l align-ment and electrical ground connections. The dielectric of all insulation was measured and corrected if necessary before equipment was energized. The control power for operating major motor start~rs is supplied from the station batteries. These batteries are kept at a constant voltage which is monitored continuously for voltage variations or undesired ground connections. Each major motor or other piece of electrical equipment is protected by over-current relays that will disconnect the de~ice if the load current becomes excessive. The protective relays are set and calibrated by trained personnel of the applicant. The availability and proper action of standby equipment is tested periodically while the unit is in operation.

8.6-2 12-1-69 Testing of automatic operation of the voltage transfer system at the 4,160 v level is performed just before startup and after a scheduled shutdown. Successful operation of the 4,160 v transfer scheme does not prevent a unit sh.utdown but is designed to provide station auxiliary power automatically when a main generator is out of service. Each standby power system was installed and checked out several months prior to criticality. The initial installation was tested to verify the starting speed and loading ability before being accepted. After acceptance, the emergency power systems were operated.on a routine test schedule. These routine operations for several months prior to criticality recognized the initial failure rate and was sufficient to achieve a proven and mature standby power system. Automatic starting and loading of the emergency generators are an essential part of the Engineered Safeguards and are arranged for periodic testing. To conduct this test, all operating safeguards equipment is connected to an emergency bus that is not to be tested. The alternate emergency bus is then given full operational test by opening its normal source breaker. The loss of voltage on the bus unde; test automatically starts the emergency generator, closes the generator breaker, and re-energizes that emergency bus system. By placing the starters in. either the operating or test position, individual components or systems may be checked completely or the test limited to the operation of the motor starters. During the testing of one emergency generator system, the alternate system is still available if required.

-* SF 8.6-3 3-15-"}1 A preventive maintenance program is followed by the Applicant to test period-ically the insulation values of circuits and equipment. During power operation, the station batteries are periodically checked for specific gravity and individual cell voltages. An equalizing or overvoltage charge is applied long enough to bring all cells up to' an equal voltage. Over a period of time, these tests reveal a weak or weakening trend in any cell, and replacements are made, if necessary. A disconnected battery or broken cell connector would be revealed during these equalizing charges. Periodically, the battery charger is disconnected and the ability of the battery,to maintain voltage and assume the direct current load is verified. This test will uncover

,* any high resistance connections or cell internal malfunctions.

During construction, checks and inspections were made to ensure that complete separation was maintained between vital equipment to ascertain redundant systems. The separation of the d-c power supply system will be verified prior to operation by performing functional checks on the two battery trains. Verification will be provided by removing one battery train from service and operating equipment on the other train. Checks will be made to ascertain that the proper equipment was actuated. This procedure will be followed for checking both d-c battery trains .

9-i 12-1-69 TABLE OF CONTENTS Section Title 9 AUXILIARY AND EMERGENCY SYSTEMS 9.1 CHEMICAL AND VOLUME CONTROL SYSTEM 9.1-.:!. 9.1.1 DESIGN BASES 9.1.1-1 9.1.2 SYSTEM DESIGN AND OPERATION 9.1.2-1 _I 9 .1.3 SYSTEM DESIGN EVALUATION 9.1.3-1 9.1.4 MINIMUM OPERATING CONDITIONS 9 .1.4-1 9 .1.5 TESTS AND INSPECTIONS 9.1.5-1 9.2 BORON RECOVERY SYSTEM 9.2.1-1 9.2.1 DESIGN BASES 9.2.1-1 9.

2.2 DESCRIPTION

9.2.2-1

- 9.2.3 9.2.3.l 9.2.3.2 DESIGN EVALUATION System Reliability Malfunction Analysis 9.2.3-1 9.2.3-1 9.2.3-2 9.3 RESIDUAL HEAT REMOVAL SYSTEM 9.3.1-1 9.3.1 DESIGN BASES 9.3.1-1 9.3.2 SYSTEM DESIGN AND OPERATION 9.3.2-1 9.3.3 SYSTEM DESIGN EVALUATION 9.3.3-1 9.3.4 TESTS AND INSPECTIONS 9.3.4-1 9.4 COMPONENT COOLING SYSTEM 9.4-1 9.4.1 DESIGN BASES 9.4.1-1 9.4.1.1 Component Cooling Water Subsystem 9.4;1-l 9.4.1.2 Chilled Component Cooling Water Subsystem 9.4.1-2 9:4.l.3 Chilled Water Subsystem 9.4.1-3

9-ii 12-1-69 Section 9.4.1.4 9.4.1.5 Title Neutron Shield Tank Cooling Water Subsystem Charging Pump Cooling Water Subsystem 9.4.1-4 9.4.1-4 9.4.2 PIPING AND VALVES 9.4.2-1 9.

4.3 DESCRIPTION

9.4.3-I 9.4.3.1 Component Cooling Water Subsystem 9.4.3-1 9.4.3.2 Chilled Component Cooling Water Subsystem 9. 4. 3-5, 9.4.3.3 Chilled Water Subsystem 9.4.3-6 9.4.3.4 Neutron Shield Tank Cooling Water Subsystem 9.4.3-7 9.4.3.5 Charging Pump Cooling Water Subsystem 9.4.3-8 9.4.4 DESIGN EVALUATION 9.4.4-1 9.4.4.1 Component Cooling Water 9.4.4-1 9.5 FUEL PIT COOLING SYSTEM 9.5-1 9.5.1 DESIGN BASES 9.5-1 9.

5.2 DESCRIPTION

9.5-8 9.5.2.1 Components 9.5-8 9.5.3 DESIGN EVALUATION 9.5-9 9.5.3.1 Availability and Reliability 9.5-9 9.5.3.2 Purification of Water 9.5-9 9.5.3.3 Storage Pit Water Leakage Control 9.5-9 9.6 SAMPLING SYSTEM 9.6.1-1 9.6.1 DESIGN BASES 9.6.1-1 9.

6.2 DESCRIPTION

9.6.2-1 9.6.3 DESIGN EVALUATION 9.6.3-1 9.7 VENT AND DRAIN SYSTEM 9.7.1-1 9 o { 1 el DESIGN BASES 9.7.1-1

9-iii 12-1-69

Section 9.7.2 9.7.3 Title DESCRIPTION DESIGN EVALUATION 9.7.2-1 9.7.3-1 9.7.3.1 System Reliability 9.7.3-1 9.7.4 TESTS AND INSPECTIONS 9.7.4-1 9.8 COMPRESSED AIR SYSTEMS 9.8-1 9.8.1 DESIGN BASES 9.8-1 9.

8.2 DESCRIPTION

9.8-7 9.8.3 DESIGN EVALUATION 9.8-9 9.8.4 TESTS AND INSPECTIONS 9. 8-11 9.9 SERVICE WATER SYSTEM 9.9-1 9.9.1 DESIGN BASES 9.9-1 9.9.2-1 9.

9.2 DESCRIPTION

9.9.2.1 Inspection and Testing 9.9.2-2 9.9.2.2 Accident Design Basis 9.9.2-2 9.9.2.3 Emergency Service Water Pumps 9.9.2-5 9.9.3 DESIGN EVALUATION 9.9.3-1 9.9.3.1 System Reliability

9.9.3-2 9.9.3.2 Malfunction Analysis 9.9.3-3 9.9.4 TESTS AND INSPECTIONS 9.9.4-1 9.10 FIRE PROTECTION SYSTEM 9.10.1-1 9.10.1 DESIGN BASES 9.10.1-1 9 .

10.2 DESCRIPTION

9.10.2-1 9.10.3 DESIGN EVALUATION 9.10.3-1 9.10.4 TESTS AND INSPECTIONS 9.10.4-1 9.11 WATER SUPPLY AND TREATMENT SYSTEMS 9.11.1-1

9-iv 12-1-69 Section 9 .11.l ~ .11. 2 Title WELL WATER SYSTEM DOMESTIC WATER SUPPLY SYSTEM 9.11.1-1 9 .11. 2-1. 9 .11. 3 FLASH EVAPORATOR SYSTEM 9.11.3-1 9.12 FUEL. HANDLING SYSTEM 9.12-1 9.12.1 DESIGN BASES 9.12.1-1 9.12.2 SYSTEM DESIGN AND OPERATION 9.12.3 FUEL HANDLING STRUCTURES 9.12.3-1 9.12.4 REFUELING EQUIPMENT 9.12.4-1 9.12.5 REFUELING PROCEDURE 9.12.5-1 9.12.6 SYSTEM DESIGN EVALUATION 9.12.6-1 9.12.7 MINIMUM OPERATING CONDITIONS 9.12.7-1 9.12.8 TESTS AND INSPECTIONS 9.12.8-1 9.13 9.13.1 9.13.2 AUXILIARY VENTILATION SYSTEM GENERAL DESCRIPTION DESIGN BASES 9.13.1-1 9.13.1-1 9.13.2-1 9.13.3 SYSTEM DESCRIPTIONS 9.13.3-1 9.13.3.1 Auxiliary Building Ventilation 9.13.3-1 9.13.3.2 Fuel Building Ventilation 9.13.3-2 9.13.3.3 Decon~amination Building Ventilation 9.13.3-3 9.13.3.4 Safeguards Areas Ventilation 9.13.3-3 9.13.3.5 Service Building Ventilation 9.13.3-4 9.13.3.6 Main Control and Relay Room Area Ventilation 9.13.3-5 9.13.4 DESIGN EVALUATION 9.13.4-1 9.13.4.1 Incident Control 9 .13. 4-2 9.13.4.2 Malfunction Analysis 9.13.4-4 __J

9-v 12-1-69

Section 9.13.4.3

  , 9 .14 Title Tests and Inspections DECONTAMINATION FACILITY 9 .13. 4-4 9.14-1 9.14'.i     DESIGN BASES                    9.14.1-1 9.. 14 .2   DESCRIPTION                     9.14.2-1 9.14.3      DESIGN EVALUATION               9.14.3-1 9.14.3.1      Malfunction Analysis          9.14.3-1 9.14~.4     TESTS AND INSPECTIONS           9.14.4-1

9.1-1 4-15-70 9.1 CHEMICAL AND VOLUME CONTROL SYSTEM The Chemical and Volume Control System adjusts the concentration of

                                              ,,,.;:::::.::::::----~-      ---- ,

chemical neutron absorber for chemical(reactivity contra,.!,

                                            \                                     " ......

maintains the proper~in~:::-;;,\i.n the Reac~'~;Cooi~;-s;:tem, provides

         '-----            j the r e q u i r e d ~ ~ for the reactor coolant pump shaft seals, provides high pressure flow to the ~~io_I!_S:y:s-t'em, provides reactor~~~:~~"a..nd degasification, and maintains the proper concentration of cf;~~!_?~ inhi.b-iting ch~~i:cals in the reactor coolant.

The system also is used (t~f:~~ ;e:*~tor Coolant System. Draining of the Reactor Coolant System to the Primary Drain System is also accomplished through the excess letdown flow path of the Chemical and Volume Control System. During normal operation, this system has provision for injecting the following chemicals, as required: l, Hydrogen to the volume control tank 2, Nitrogen for purging the volume control tank

3. Lithium hydroxide via the chemical mixing tank to the charging pumps suction,

9 .1.1-1 4-15-70 a 9 .1.1 DESIGN BASES During normal unit operation, the Chemical and Volume Control System is designed t o ~ ~ y provid.e-*boric***rc;id-*::;:u~~~t a preset concentration, ( __.// .....__ __ --------"*----*-~---

                                                  ,.~                   ..,_

which matches the Reactor Coolant System boron concentration, to compensate for minor leakages of reactor coolant. The Chemical and Volume Control System design alsoi:ermits the additions of a pre-selected quantity of either reactor p~~ry grade makeup. wat~ or~~~~ated _bori~--~=~~9lut;j.ori.-;t' a pre-selected flow rate to the Reactor Coolant System. The Chemical and Volume Control System has the capacity to provide a cold shutdown for the two units, each with one control rod assembly completely withdrawn following a refueling shutdown on both units. One boric acid tank has sufficient capacity (if maintained above the low level alarm pain~) to provide a cold shutdown for one unit with one control rod assembly completely withdrawn. 9.1.1.1 Redundancy of Reactivity Control In addition to the reactivity control achieved by the control rod assemblies as detailed in Section 7.3, reactivity control is provided by the Chemical and Volume Control System which regulates the concentration of boric acid solution in the Reactor Coolant System. The system is designed tc, prevent,

9.1.1-2 12-1-69 under postulated system malfunction, uncontrolled or inadvertant reactivity changes which might stress the system beyond design limits. 9.1.1.2 Reactivity Shutdown Capability Normal reactivity shutdown capability is provided by control rods with boric acid injection used to compensate for the xenon transient and for unit cool-down. Any time that the unit is at power, the quantity of boric acid retained in the boric acid tanks and ready for injection always exceeds that quantity required for a cold shutdown. The boric acid solution is transferred from the boric acid tanks by boric acid transfer pumps to the suction of the charging pumps, which inject boric acid into the reactor coolant. Any charging pump and any boric acid transfer pump is capable of being operated from diesel generator power on loss of primary power. Boric acid is injected by one charging and one boric acid transfer pump at the approximate reactivi5l __insertion rate of -0 .24%

      .                   (_,--    ~
                                                            ~t per minute, which shuts the reactor down irl~s'fifteen
                           ,,*,       'minutes
                                       \       with no rods inserted. In fifteen
                             *,~/

additional minutes, enough boric acid can be injected to compensate for xenon decay although xenon decay below the equilibrium operating level does not begin until approximately 12-15 hours after shutdown from full power. If two'boric acid transfer pumps are available, shutdown periods are halved. Additional boric acid is added if it is desired to bring the reactor to cold shutdown conditions. e

9.1.1-3 12-1-69 On the basis of the above, the injection of boric acid provides backup shut-down reactivity capability, independent of control rod assemblies which normally serve this function in the short term situation. Shutdown for long term and reduced temperature conditions is accomplished with boric acid injection using redundant components. The reactivity control systems provided are capable of making and holding the core subcritical for any cold shutdown, hot shutdown, or hot operating condition, including those resulting from power changes. The maximum excess reactivity expected for the core occurs for the cold, clean condition .at the beginning of life. A total of 48 full length and 5 part length control rod assemblies is provided. The 48 full length assemblies are divided

into two categories comprising control and shutdown groups.

The control group, used *in combination with soluble boron, provides control of the reactivity changes of the core throughout the life of the core at power conditions. This group of *control rods assemblies is used to compensate for short term reactivity changes at power that might be produced due to variations in reactor power requirements or in coolant temperature. The soluble boron control is used to compensate for the slower changes in reactivity throughout core life such as those due to fuel depletion and fission product buildup and decay.

The reactor core, together with the Reactor Control and Protection System, is designed so that the minimum allowable DNBR will be at least 1.30 and there is no fuel melting during normal operation including anticipated transients. Shutdown groups are provided to supplement the control group of control rod assemblies to make the reactor at least one percent subcritical (keff = 0.99) following trip from any credible oper~ting condition to the hot shutdown, condition. This assumes the highest worth control rod assembly remains in the fully withdrawn position. Sufficient shutdown capability also is provided to maintain the core sub-critical for the most severe anticipated cooldown transient associated with a single active failure, e.g., accidental opening of a steam bypass or relief valve. This is achieved with a combination of control rod assemblies and automatic boron addition via the Safety Injection System with the highest worth rod being fully withdrawn. Manually controlled. boric acid addition is used to maintain the shutdown margin for the long term conditions of xenon decay and reactor system cooldown. 9.1.1.3 Codes and Classifications The codes and classifications of pressure retaining components through which reactor coolant circulates at reduced pressures and temperatures, or which

9.1.1-5 12-1-69 - are exposed to reactor operating pressures and temperatures, are stated in Table 9, l-1. Both the regenerative and excess letdown heat exhangers are classified Class C according to ASME Boiler and Pressure Vessel Code, Section III. These heat exhangers meet all of the requirements for Class C vessels. Westinghouse supplements these minimum requirements with the following additional require-ments:

1. Wel.ded tube to tube sheet joints.

2. Gas leak test of tube to tube sheet welds in addition to full different-ial pressure hydrostatic tests *

3. Special tube to tube sheet weld procedure qualifications.

4. Ultrasonic or eddy current test of tubing.

5. Dye penetrant examination of tube to tube sheet welds and root pass as well as final pass to all other pressure containing welds.

6. Fatigue analysis as required by paragraph 415.1 of Section III to demonstrate that the unit can withstand the transients that it is expected to experience during its design life

In addition, present Westinghouse equipment specifications for the regenerati'.v.e and excess letdown heat exchangers meet the basic requirement of Appendix e:

I IX of .the ASME Code except that Mes tinghouse does not require non-destructive test personnel to *be qualified to American Society for Non-destructive Tes ting procedures. Where the suppliers' personnel *are not so qualified, Westinghouse assures that suppliers' personnel are adequately qualified by periodic obse.rvation of their performance and also performs the customary final inspections. As noted above, Wes:tinghouse Quality Assurance levels and Quality Control procedures are in excess of standard code requirements -for Class C vessels. Austenitic stainless steel or other corrosion resistant materials are used for surfaces or components in contact with either reactor coolant o;r boric acid solution.

9.1.1-7 12-1-69 TABLE 9.1-1 CHEMICAL AND VOLUME CONTROL SYSTEM CODE REQUIREMENTS Regenerative heat exchanger ASME III*, Class C Non-Regenerative heat exchanger ASME III, Class C, Tube Side ASME VIII, Shell Side Mixed bed demineralizers ASME III, Class C Reactor coolant filter ASME III, Class C Volume control tank .ASME III, Class C Seal water heat exchanger ASME III, Class c, Tube Side ASME VIII, Shell Side Excess letdown heat exchanger ASME III, Class C, Tube Side ASME VIII, Shell Side Chemical mixing tank ASME VIII Cation bed demineralizer ASME III, Class C Boric acid tanks ASME VIII Deborating demineralizer ASME III, Class C

 , Batching .tank                                     ASME VIII Seal water injection filters                       ASME III, Class C Pumps                                              No Code Boric acid filter                                  ASME III, Class C Seal water filter                                  ASME III, Class C Resin fill tank                                    No Code Piping and valves                                  USAS B31.l** and USAS B16.5***

* ** *** Notes on following page.

9.1.1-8 12-1-69

ASME III - American Society of Mechanical Engineers, Boilers and Pressure Vessel Code, Section III, Nuclear Vessels.

i

- USAS B31.l - Code for Pressure Piping, American Standards Association, and special nuclear cases where applicable.
  - USAS Bl6.S - Code for Steel Pipe Flanges and Flanged Fittings, American Standards Association.

9.1.2-1 12-1-69 9 .1.2 SYSTEM DESIGN AND OPERATION The Chemical and Volume Control System shown in Figures 9.1-1 thru 9.1-6 provides a means for injection of negative reactivity in the form of boric acid solution, chemical additions for corrosion control, and reactor coolant cleanup and degasification. This system also adds makeup water to the Reactor Coolant System, reprocesses the water letdown from the Reactor Coolant System, provides seal water injection to the reactor coolant pump seals, provides high pressure flow to the Safety Injection System, and fills the Reactor Coolant System. Draining of the Reactor Coolant System to the Primary Drain System is also accomplished through the excess letdown flow path of the Chemical and Volume Control System *

The system is provided with overpressure devices, such as safety valves, to protect components whose design pressure and temperature are less than the Reactor Coolant System design limits. System discharge from overpressure protective devices and other system leakages are directed to closed systems.

System design enables post-operational* hydrostatic testing to applicable code test pressures. The relief valves are gagged during hydrostatic testing. The relief valves in systems that are hydrostatically tested after refueling operations are set at the system design pressure. Five tanks in the Chemical and Volume Control System are shared by the two units. These tanks are listed in Table 9.1-3 .

9.1.2-2 12-1-69 9.1.2.1 System Description During normal unit operation, reactor coolant flows through the letdown line from the reactor coolant pump discharge side of reactor coolant loop number 1 cold leg, and returns through the charging line to the reactor coolant pump discharge side of the cold leg of loop number 2. An alternate letdown path from the Reactor Coolant System is provided in the event that the normal letdoun oa+-h is inopPrable. Reactor coolant can be discharged from each reactor coolant loop, or all loops concurrently, through the common loop drain header to the tube side of the excess letdown heat exchanger. Each of the connections to the Reactor Coolant System loops has an isolation valve located close to the loop piping. The charging line has a check valve located dowstream of the charging line isolation valves. Reactor coolant entering the Chemical and Volume Control System flows through the shell side of the regenerative heat exchanger where its temperature is reduced. The coolant then flows through a letdown orifice which reduces the coolant pressure. The letdown flow leaves the reactor containment and enters the auxiliary building where it undergoes a second temperature reduction in the tube side of the non-regenerative heat exchanger, followed by a second pressure reduction by a low pressure letdown valve. After passing through one of the mixed bed demineralizers, where anionic and cationic impurities are removed, coolant flows through the reactor coolant filter and enters the volume control tank through a spray nozzle. Reactor coolant letdown flow is diverted to the Boron Recovery System (Section 9.2) on a high level signal from the volume control tank.

9.1.2-3 12-1-6~

The cation bed demineralizer, located downstream of the mixed bed deminera-lizer, is used intermittently to control cesium activity in the coolant and also to remove excess lithium which is formed from BlO (n, a) Li 7 reaction.

Hydrogen is automatically supplied, as determined by pressure control, to the vapor space in the volume control tank, which is predominantly hydrogen and water vapor. The hydrogen within this tank is, in turn, the supply source to the reactor coolant. Fission gases are periodically removed from the system by venting the volume control tank to the Vent and Drain System (Section 9.7) or by diverting the letdown stream to the gas stripper in the Boron Recovery System prior to a cold or refueling shutdown~ The coolant flows from the volume control tank to the charging

pumps which ra~se the coolant pressure above that in the Reactor Coolant System. The coolant then enters the containment, passes through and is heated in the tube side of the regenerative heat exchanger, and then returns to the Reactor Coolant System.

A portion of the high pressure charging flow is injected into the reactor coolant pumps between the pump impeller and the shaft seal so that the seals are not exposed to high temperature reactor coolant. Part of the flow cools the lower radial bearing and enters the Reactor Coolant System tlirough a labyrinth seal on the pump shaft. The remainder which is the shaft seal leakage flow, is filtered, cooled in the seal water heat exchanger. and returned to the suction of the charging pumps. Coolant injected through the reactor coolant pump labyrinth seals returns to the volume control

tank by the normal letdown flow path through the regenerative heat exchanger.

9.1.2-4 12-1-69 When the normal letdown flow route is not in service, labyrinth seal injection flow is returned to the suction of the charging pumps through the excess letdown and seal water heat exchangers. Boric acid is dissolved in heated water in the batching tank to a concentration of approximately 12 weight percent. The lower portion of the batching tank is jacketed to utilize low pressure steam to permit heating of the batching tank solution. One of four boric acid transfer pumps is used to transfer this concentrated solution to the boric acid tanks. Small quantities of boric acid solution from the boric acid tanks are metered from the discharge of an operating boric acid transfer pump for blending with the water supplied to makeup for normal leakage losses, or for increasing the reactor coolant boron concentration during normal load follow operation. Electric immersion heaters maintain the temperature of the solution in the boric acid tanks at approximately 165°F to prevent precipitation. During unit startup, normal operation, load reductions and shutdowns, liquid effluents ~ontaining boric acid flow from the Reactor Coolant System through the letdown line and are collected in the Boron Recovery System where the boric acid is recovered and stored (Section 9.2). Cover gases displaced during the filling of volume control tanks are vented to the Gaseous Waste Disposal System. During the unit cooldown phase when the residual heat removal loop is operating, and the letdown orifices are not in service, a flow path is provided to remove fission products, corrosion and other solid and

9.1.2-5 12-1-69 liquid impurities. A portion of the flow leaving the residual heat exchangers passes through the non-regenerative heat exchanger, mixed bed demineralizers, reactor coolant filter and volume control tank. The fluid then is pumped by the charging pump through the tube side of the regenerative heat exchanger into the Reactor Coolant System and, through the auxiliary spray line, into the pressurizer. Tables 9.1-2, 9.1-3 and 9.1-5 list the system performance requirements, data for individual system components and reactor coolant equilibrium activity concentration. 9.1.2.2 Reactor Coolant Activity Concentration

The parameters used in the calculation of the reactor coolant fission product inventory, including pertinent information concerning the coolant cleanup flow rate and the demineralizer effectiveness, are presented in Table 9.1-4. The results of the calculations are presented in Table 9.1-5. In these calculations, the defective fuel rods are assumed to be uniformly distributed throughout the core and the fission product escape rate coefficients are therefore based upon an average fuel temperature.

Volume control tank noble gas concentrations with 1% failed fuel are shown in Table 9.1-6

9.1. 2-6 12-:).,069 TABLE 9.1-2 CHEMICAL AND VOLUME CONTROL SYSTEM PERFORMANCE REQUIREMENTS*

Station design life, years 40 Nominal pump seal water supply flow rate (to Reactor Coolant Pumps), gpm 24 Nominal pump seal water return flow rate (from Reactor Coolant Pumps), gpm 9 Normal letdown flow rate, gpm 60 Maximum design letdown flow rate, gpm 120 Normal charging pump flow rate (one pump including 60 gpm bypass flow) gpm Normal charging line flow, gpm Maximum rate of boration with one transfer and one 129 45

charging pump, from initial Reactor Coolant System concentration of 1800 ppm, ppm per min 30 Equivalent cooldown rate during the above rate of boration, °F per min 8.8 Maximum rate of boron dilution with maximum design letdown flow rate from initial Reactor Coolant System concentration of 2500 ppm, ppm per hr 350 Two-pump rate of boration using refueling water from initial Reactor Coolant System concentration of 10 ppm, ppm per min 14.6
Volumetric flow rates in gpm are based on 130°F and 2350 psig.
9.1.2-7 12-1-69

- TABLE 9.1-2 (Continued) Equivalent cooldown rate during the two-pump rate of boration, °F per min 4.1 Temperature of reactor coolant entering system at full power with normal letd.own and charging line flow rates, °F 543.5 Temperature of reactor coolant return to Reactor Coolant System at full power, °F , 489 .3 Normal system discharge temperature to Boron Recovery System, °F J,.15 Approximate amount of 11.5 percent boric acid solution required to meet cold shutdown conditions, gal 2700*

Range of boric acid concentration is 11.5 to 12.5 percent. The amount of solution is determined from the lower limit of concentration in order to obtain the more conservative figure
TABi,E 9.1-3 CHEMICAL ANp VOL~ CONTROL_ SYSTEM PRINCIPAL COMPONENT DATA

SUMMARY

Design t1aximum Design

                                }J.eat              i,etdown  Letdown    D~sign             Temper-Quantity    Transfer
                                     . .   ,        Flow,    L':I T,     fr~sS!ure, psig  .a,ture~ OF
                    ]2er Unit   Btuihr              lb/hr     OF          shell/tube     .. shell/tube Heat Exchangers Regenerative    1            8.34   X 10 6       29,826   260.5       2485/2735         650/650 (norm) 6 15.2    X 10         59, 700  239         f485/2735         650/650 (max)

Nonregenerative 1 15.8 X 10 6 59,700 265 150/600 250/400

S~al Water (Floating Ring 6 Seals) 1 1.09 X 10 111,600 26 150/150 250/250 lj\O

                                                                                                       ....II *N~

Excess Letdown 1 3.23 X 10 6 7,500 408 150/2485 250/650 C'I ..

                                                                                                       \0 ex>

TABLE 9.1-3 (Continued) Design Design Quantity Capacity, Head, ft Pressure , Temper-per Unit gpm or psig psig ature, °F Pumps Charging 3 Centrifugal 150'** 5800 ft ** 2735 250 Boric Acid transfer 2 Canned 75 235 ft 150 250 Design Design Quantity Pressure, Temper-per Unit Volume psig ature, OF

Tanks 3 75 Int/15 Ext 250 Volume control 1 Vertical 300 ft Chemical mixing 1 Vertical 5 gal 150 250 I-' \0 Boric acid 3* Vertical 7500 gal Atmos. 250 N, I I-' I-'. IN Batching 1* Jacket Btm. 800 gal Atmos. 250 °'

                                                                                                     '°   I
                                                                                                         \0 Resin fill          1*          Vertical         635 gal                 Atmos.         200
*Shared lly both units    ** Charging mode

TABLE 9.1-3 (Continued) Resin Design Design Design Quantity Volume, Flow, Pressure, Temper-gpm psig ature, °F Demineralizers Mixed bed 2 Flushable 30 120 200 250 Cation bed 1 Flushable 20 60 200 250 Deborating 1 Fixed 43 120 200 250 Design Design Design Particle Quantity Flow, Temperature, Pressure. Retention (with 98% efficiency), 12er Unit gEm OF psig microns Filters Reactor coolant 1 JOO 250 200 25 Seal.water 1 300 250 200 25 I-' \C Boric acid 1 300 250 200 25 No I ._. I-'

IN Seal water CJ\ I

                                                                                          \CI-'

injection 2 80 200 2735 5 0

9.1.2".""ll 12-1-69

- The f1ssion product c1ctivi ty in the reactor coolant in th~ letdown stream of the regenerative heat exchanger during opeiation ~ith small cladding defects in 1% of the .fuel rods is computed using the following differential equations:

For parent nuclides in the coolant dN. W1 dt = Dvi NC. B' ) N . W1 1 B - tB' 0 for daughter nuclides in the coolant, dN wj dt Dv. NC. (X. 1

                                  + Rn.1 +  ~~~~~~~

B' ) NW]. + X.N . 1 W1 J B - tB' J 0 where:

N D

                = population of nuclide units
                = fraction of fuel rods having defective cladding R      = purification flow, coolant system volumes per sec.

B 0

                = initial boron concentration, ppm B'     = boron concentration reduction rate by feed and bleed, ppm per sec.

t = time, seconds or fraction n = removal efficiency of purification cycle for nuclide X = radioactive decay constant u = escape rate coefficient for diffusion into coolant

9.1.2-12 10-15-70 Subscript C refers to core Subscript w refers to coolant Subscript i refers to parent nuclide Subscript j refers to daughter nuclide 9.1.2.2.1 Tritium Production Within a Light Water Reactor A. General - Overall Sources Within a pressurized light water reactor, tritium is formed from several sources. The greatest potential source is the fissioning of uranium fuel, which yields tritium as a ternary fission product at a rate of approximately

      -5                                  -2 8 x 10     atoms per fission, or 1.05 x 10    curies/mwt/day. Boron-bearing control rods will also be a source of tritium. The amount of tritium appear-ing in the reactor coolant from these two sources is a function of the fuel and control rod cladding material permeability to tritium.

A direct source of tritium in the reactor coolant is the reaction of neutrons with dissolved boron used for reactivity control. The boron concentation is approximately 1000 ppm at the beginning of the fuel cycle and is reduced to zero at the end of the fuel cycle. Neutron reactions with lithium are also a direct source of tritium. Lithium is present for pH control, and as a product of boron reactions with neutrons. The amount of lithium present, however, is carefully controlled to approximately 2.2 ppm by demineralization. A minute amount of tritium is also produced by neutron reactions with naturally occurring deuterium in light water.

9 .1. 2.,..,12a 10-15-70

- B.

1. Specific Individual Sources of Tritium - Light Water Reactors Ternary Fissions - Cla<l Diffusiop I I A program has been undertaken by Westinghouse to determine the source of I tritium in the reactor .coolant in operating plants with both stainless steel I I and zircaloy cladding. Th;i.s program has clearly indicated that with the current generation of Westingho4se reactors with zircaloy clad fuel, 1% or II

                                                                                  !I less of the tritium produced in the fuel will diffuse through the cladding into the coolant.

The Ginna plant operates at 1455 MWt and has z;i.rcaloy cladding. After approximately 8 months of operation, the tritiu~ concentrations are less

than .3µCi/cc in the reactor coolant.

have averaged parable.

                ~s  curies/month.

The monthly discharges from the plant The experience at NOK and Zorita are com-An extensive program to follow the buildup of tritium in the Ginna station is underway, and the results to date indicate a potential source from the core whiGh is 1.% or less of the ternary fissions generated in the fuel. Westinghouse has in the past assumed that 30% of the tritium from ternary fissions would diffuse through the zircaloy fuel. This value was used as a basis for syste~s and operational design. Present experience indicates that this was conservative .

9 .l.2-12b 10-15-70

2. Tritium Produced from Boron Reactions the neutron reactions with boron resulting in the production of tritium are:

BlO (n' 2 ex:) T BlO (n, ex:) L/ (n, n ex:) T Bll (n? T) Be 9 BlO (n, ci) Be 9* (n, ex:) Li 6 (n, ex:) T Of the above reactions, only the first two contribute significantly to the 11 9 tritium production in a PWR. The B (n, T) Be reaction has a threshold of 14 Mev and a cross section of 1\.,5 mb. Since the number of neutrons produced 2 at this energy are less than 10 9 n/cm -sec,: the tritium produced from . this reaction is negligible. since the Be 9* The B 10 (n, d) reaction may be neglected produced in this reaction has been found to be unstable. 3, Tritium Produced from Lithium Reactions The neutron reactions with lithium resulting in the production of tritium are: L / (n, n ex:) T 6 Li (n, ex:) T .I In Westinghouse design reactors, lithium is used for pH adjustment of the I I reactor coolant. The reactor coolant is maintained at a maximum level of

9 .1.2-12c 10-15-70 7 2.2 ppm lithium by the addition of Li 0H and by a cation demineralizer included in the Chemical and Volume Control System. This demineralizer will remove any excess of lithium such as could be produced in the BlO I 7 (n, ~) Li reaction. I I

i 6 6 The Li (n, ~) T reactor is controlled by limiting the Li impurity in ! I 7 the Li 0H used in the reactor coolant and by lithiating the demineralizers 7 with 99.9 atom percent Li .

4. Control Rod Sources In a fixed burnable poison rod, there are two primary sources of tritium generation includi.ng the BlO (n, 2a) T and the BlO (n, a) 1/ (n, na)

T reactions. Unlike t;he coolant where the L/ level is controlled at 2ppm, there is a buildup of Li 7

in the burnable poison rod. poison rods are required only during the first core cycle. 10 The burnable During this time the tritium production is 72 curies/pound B . The control rod materials used are Ag-In-Cd which are not tritium sources. 5, Tritium Production from Deuterium Reactions Since the amount of naturally occurring deuterium in water is less than o.ooi5, the tritium produced from this reaction is negligible (less than 1 curie per year) .

9 .1. 2-12d 10-15-70

6. Total Tritium Sources Tritium sources in the reactor coolant system of the Surry Urtit are listed in Tables 9.1-7 and 9.l-*7A, which are presented on the basis of 12 months of operation at full power at a 0.8 load factor.

Two columns are presented in the tables - a previous design value and the presently expected tritium release value to the reactor coolant. The design values are based on a release of 30% of the tritium produced being diffused through the fuel cladding. 9.2.2.2 Activity Monitoring and Control During unit operation, continuous monitoring of the reactor coolant.is accomplished by means of high range and low range gross activity monitors . These monitors, which are described in Section 11.3.3, are capable of determining any sudden increase in activity level due to failed fuel within the range of 10-4

                    ,µCi/cc to 10 3
                                    µCi/cc.

The Technical Specification limit on reactor coolant activity can be shown to provide adequate protection to the general public as follows. The limits on Reactor Coolant System leakage and on effluent release govern the potential release of coolant activity to the environment during normal reactor operation, and have been established.on the basis of the limiting values of reactor coolant activity. The limitation on containment leak rate provides protection for breaks in the Reactor Coolant System that exhaust to the containment. I:t remains to consider breaks into other closed systems. The reference ~ccident considered for the bases is the steam generator tube rupture (See Section 14.3.1).

9.1.2-13 10-15-70

Rupture of a steam generator tube would allow a portion of the reactor coolant activity to -enter the Steam and Feedwater System outside the containment. In this event, the radioactive noncondensable gasses would be detected by the radiation monitor located in the air ejector effluent line. When the radioactivity level reaches the alarm set point of- the monitor, trip valves in the effluent line will automatically actuate to divert the flow to the containment and to close the vent to atmosphere.

I The steam generator tube rupture is considered to be a highly improbable accident and in rio case would this accident result in exceeding the guide-lines of 10 CFR 100. A tritium limit is establi~hed to meet the allowable concentration in the circulating water discharge. For one unit the production rate of

tritium due to ternary fission is calculated to be 7850 curies/year and 30%, as a design basis, is assumed to be released to the coolant by recoil through the cladding (See Table 9.1-7). To this is added tritium from other sources for a total of approximately 2745' ~~rie~/ye~;. ;f total tritium acitivty added to the reactor coolant during the initial fuel cycle and 2750 curies/year during the equilibrium fuel cycle. Using the projected turnover rate of four Reactor Coolant System volumes per year or more~

the tritium activity in the primary coolant_will never increase beyond* about 2.5 µCi/cc .

9.1.2-14 12-1-69 9.1.2.3 Reactor Makeup Control Modes The reactor makeup control consists of an instrument and control group arranged to provide a manually pre-selected makeup composition to the charging pump suction header or the volume control tank. The makeup control functions are designed to maintain desired operating fluid inventory in the volume control tank and to adjust reactor coolant boron concentration for reactivity and shim control. Makeup for normal primary system leakage is regulated by the reactor makeup control which is set by the operator to blend water from the primary water tanks with concentrated boric acid to match the reactor coolant boron concentration. Makeup is added automatically if the volume control tank level falls below a preset value. The reactor makeup control is designed to operate from the Main Control Room by manually pre-selecting ma~eup composition to the charging pump suction header or the volume control tank. This maintains the desired operating fluid inventory in the volume control tank and adjusts the reactor coolant boron concentration for proper reactivity control. The operator can stop the makeup operation at any time in any operating mode by remotely closing the makeup stop valve*s.

9.1.2-15 12-1-69

One primary water supply pump and one boric acid transfer pump normally are operated. If either pump trips, an alarm due to deviation of flow rate from control set point alerts the operator and a standby pump is started manually
9.1..2:-16 12-1-69 TABLE 9.1-4 P~ETERS- USED. IN THE CALCULATION OF REACTOR COOLANT FISSION PRODUCT ACTIVITIES

1. Core thermal power, *maximum expected rating, MWt 2546

2. Fraction of fuel containing clad defects 0.01

3. Reactor Coolant System liquid volume, ft 3 (including pressurizer at normal level) 9235

4. Reactor coolant average temperature, °F 560

5. Letdown purification flow rate (normal), gpm

  /*                                                     60

6. Elrective cation demineralizer flow, gpm 6

7. Volume control tank volume, ft 3

a. Vapor 180

b. Liquid 120

8. Fission product escape rate coefficients, sec -1

a. Noble gas isotopes 6.5 X 10-S

b. Br, I and Cs isotopes 1.3 X 10-8 c-. Te isotopes 1.0 X 10- 9

d. Mo isotopes 2.0 X 10- 9

e. St and Ba isotopes 1.0 X 10-11

f. Y, La, Ce and Pr isotopes 1.6 X 10-12

9.1.2-17 12-1-69

TABLE 9.1-4 (Continued)

9. Mixed bed demineralizer decontarnitiation factors:

a. Noble gases and Cs-134, 136, 137, Y-90 and Mo-99 1.0

b. All other isotopes (except Tritium) 10.0

10. Cation bed demineralizer decontamination factor for Cs-134, 136, 137, Y-90 and Mo-99 10.0

11. Initial boron concentration, ppm 1000 (equilibrium cycle, hot full power)

12.

13. Boron dilution rate, ppm per full power day Volume control tank noble gas stripping 3.46 fraction (closed system): Isotope Stripping Fraction Kr-85 2.3 X 10-S Kr-85m 2.7 X 10-l Kr-87 6.0 X 10-1 Kr-88 4.3 X 10-l Xe-133 1.6 X 10- 2 Xe-133m 3.7 X 10-2 Xe-135 1. 8 X 10-1 Xe-135m 8.0 X 10-l Xe-138 1.0

9.L2-*1s 12-1-69 TABLE 9 ~l-5 FISSION PRODUCT CONCENTRATIOO"S IN THE REACTOR COOLANT WITH SMALL CLADDING DEFECTS IN ON*E PERCENT OF THE FUEL RODS (Assumptions are stated in Table 9.1-4) Fission Product Reactor Coolant Activity Concentratiom., Isotope µCi per cc at 560°F A. Noble Gases Kr-85 2.42 (P~ak) Kr-85m 1.14 Kr-87 Kr-88 Xe-133 0.78 2.81 1.88 X 10 2 Xe-133m 1.87 Xe-135 5.20 Xe-135m 1. 30 X 10-l

                                                        -1 Xe-138                                     3.5 X 10 Sub Total 202.7 µCi per cc B. Non-Gaseous 2

Br-84 3.0 X 10-Rb-88 2.82 Rb-89 6.5 X 10-2

9.1.2-19 12-1-69

Fission Product Isotope TABLE 9.1-5 (Continued Reactor Coolant Activity Concentration,

                                 µCi per cc at 560°F Sr-89                         2.8   X   10- 3 Sr-90                         8.5   X   10- 5 Y-90                          1.0   X   10- 4 Sr-91                         1.3   X   10- 3 Y-91                          4.9   X   10- 4 Sr-92                         5.2   X   10- 4 Y-92                          5.3   X   10- 4 Zr-95                         5.4   X   10- 4 Nb-95                         5.4   X   10- 4

Mo-99 Te-129 I-129 2.23 4.6 2.4 X

X 10- 3 10- 8 I-131 1. 68 Te-132 1. 86 X 10-l I-132 6.25 X 10-l I-133 2.73 Te-134 2.14 X 10- 2 I-134 3.8 X 10-l I-135 1.43 Cs-134 1. 76 X 10-l Cs-136 2.6 X 10- 2 Cs-137 9.75 X 10-l

Cs-138 4.58 X 10- 2

9.1.2-20 12-1-69 Fission Product TABLE 9.1-5 (Continued) Reactor Coolant Activity Concentration, Isotope µCi per cc at 560°F _3 Ba-140 1.6 X lQ La-140 6 ,2 X 10- 4

                                           ..:.4 Ce-144                         2.1  X 10

_4 Pr-144 2. 3 X 10 Subtotal 13.43 µCi per cc Total 216.13 µCi per c:c

9 .1. 2-21 12-1-:-69

TABLE 9.1-6 MAXIMUM VOLUME CONTROL TANK NOBLE GAS CONCENTRATION IN VAPOR PHASE WITH SMALL CLADDING pEFECTS IN ONE PERCENT OF.THE FUEL RODS Vapor Phase Activity Concentration Isotope µCi per cc Kr-85 1.84 Kr-85m 36.3 Kr-87 7.30 Kr-88 38.6
Xe-133 Xe-133m Xe-135 3020.0 32.8 67.8 Xe-135m 0.21 Xe-138 0.72 Total 3206 µCi per cc

9.1.2-22 10-15-70 TABLE 9.1-7 TRITIUM SOURCES FROM THE REACTOR EMPLOYING AG-IN-CD ABSORBER RODS

Basic Assumptions and Plant Parameters:

1. Core thermal power 2546 MWt

2. Plant load factor 0.8

3. Core volume 3 938.7 ft

4. Core volume fractions

a. 0.2990

b. 0.0933

c. 0 .6077

5. Initial reactor coolant boron level

a. Initial cycle 720 ppm

b. Eqµilibrium cycle 1000 ppm

6. Reactor coolant volume 3 9235 £t 7.

8. Reactor coolant peak lithium level

                   . 7 (99% pure Li)

Core averaged neutron fluxes:

a. E > 6 Mev 2.2 ppm 2

n/cm -sec

                                                       .2.99   X 10 12 12

b. E>5Mev 8 .00 X 10

c. 3 Mev < E < 6 Mev 13 2.32 X 10

d. 1 Mev < E < 5 Mev 13 5 .47 X 10

e. 0.625 ev 13 E < 2 .32 X 10

9. Neutron reaction cross-sections

a. B 10 (n,2 cr) T: o (1 Mev < E < 5 Mev) = 31.6 mb (spectrum weighted) o (E > 5 Mev) = 75 mb 7

b. Li (n, ncr) T: o (3 Mev < E < 6 Mev) 39.1 mb (spectrum weighted) o (E > 6 Mev) = 400 mb

10. Fraytion of ternary tritium diffusing through zirconium cladding a.

b. Design value Expect~d value 0.30 0.01

9.1..i-2@ 2.-15-71 TABLE.9.1-7A

TRITIUM SOURCES IN REACTOR COOLANT OPERATION AT A POWER LEVEL OF 2546 MWTH CURIES/12 FULL POWER MONTHS ATAN 0.8 LOAD FACTOR (CURIES/YEAR)

Released to the Coolant Total Design Expected Tritium Source Produced Value Value Ternary Fissions 7850 2355 78.'i

               .       (1)

Burnable Poison Rods 350 105 105 (Initial Cycle) Control Rods 0 0 0 Soluble Poison Boron

(Initial Cycle) ( 2 )

(Equilibrium Cycle) ( 3 ) Li-7 Reaction 270 380 9.2 270 380 9.2 270 380 9.2 LI-6 Reaction 4.6 4.6 4.6 Deuterium Reaction 1 1 1 Totals Initial Cycle 8490 2745 468 Totals Equilibrium Cycle 8250 2750 473 (1) Weight of B 0 = 85# (BlO - 5.23#) 2 3 (2) Initial boron (hot, full power, equilibrium xenon)= 700 ppm (3) Initial boron (hot, full power, equilibrium xenon)= 1000 ppm

9 .1. 2-24 10-15-70 I I [INTENTIONALLY DELETED] i *.

9.1.2-25 10-15-70 [INTENTIONALLY DELETED]

9 .1. 2-26 12-1-69 Makeup wate'r to the Reactor Coolant Sys Lem Is provldl*d throu~h the Chemical and Volume Control System from the following sources:

1. The primary water tanks, which provide water for primary coolant dilution when the reactor coolant boron concentration is to be reduced.

2. The boric acid tanks, which supply concentrated boric acid solution when reactor coolant boron concentration is to be.

increased.

3. The refueling water storage tank which supplies borated w~ter for.emergency makeup.

4. The chemical mixing tank, which is used to inject small quantities of solution when additions of a pH control chemical are necessary.

Seal water inleakage to the Reactor Coolant System requires a contin-uous letdown of reactor coolant to maintain the desired inventory. In addition, bleed and feed of reactor coolant is required for removal of impurities and adjustment of boric acid in the reactor coolant. L

9 .1.2-27 12-1-69

- - Automatic Makeup The "automatic makeup" mode of operation of the reactor coolant water makeup control provides boric acid solution at a preset concentration to match the boron concentration in the Reactor Coolant System. The automatic makeup compensates for minor leakage of reactor coolant without causing significant change in the coolant boron concentration.

Under normal unit operating conditions, the makeup control is set for automatic operation. A_ preset low ievel si&nal from the volume control tank level controller causes the automatic makeup control action to increase the speed on the normally running boric acid transfer pump, open the makeup stop valve to the charging pump suction, and open the

concentrated boric acid c;:ontrol valve and the reactor primary water makeup

   --------~control valve*. One primary water supply pump is always in operation.

The flow controllers then blend the makeup stream according to the preset concentration. Makeup addition to the charging pump suction header causes the water level in the volume control tank to rise. At a preset high level

          .point, the makeup is stopped; the reactor primary water makeup control valve closes; the boric acid transfer pump returns to low speed; the concentrated boric acid control valve closes, and the makeup stop valve to the char&ing pump suction closes.

Dilution The "dilution" mode of operation permits addition of a pre-selected quantity of reactor primary grade water makeup at a pre-selected flow rate to the Reactor Coolant System. The operator sets the makeup stop valves to the

9 .1. 2-28 12-1-69 _ volume control tank and to the charging pump si1c~ion in the closed position, selects the "dilution" mode, sets the reactor primary water makeup flow controller set point to the desired flow rate, sets the reactor primary water makeup batch integrator to the desired quantity and initiates system start. This opens the primary grade water makeup control valve which delivers primary grade water to the volume control tank. Excessive rise of the volume control tank water level is prevented by automatic actuation of a three-way diversion valve which routes the reactor coolant letdown flow to the Boron Recovery System. When the preset quantity of reactor primary water makeup is added, the batch integrater causes the reactor primary water makeup control valve to close * .Boration The "boration" mode of operation permits the addition of a pre-selected quantity of concentrated boric acid solution at a pre-selected flow rate to the Reactor Coolant System. The operator sets the makeup stop valves to the volume control tank and to the charging pump suction in the closed position, selects the "boration" mode, sets the concentrated boric acid flow controller set point to the desired flow rate, sets the concentrated boric acid batch integrater to the desired quantity, and initiates system start. This opens the makeup stop valve to the charging pumps suction, and increases the speed on the normally operating boric acid transfer pump, which delivers a 12 percent boric acid solution to the charging pump suction header. The total quantity added in most cases is so small that it has only a minor effect on the volume control tank level.

9 .1. 2-29 12-1-69

When the preset quantity of concentrated boric acid solution is added, the batch integrator causes the boric acid transfer pump to /return t:-o-* iow and* c;toefes
tne*makeup stop valve to the suction of the charging pumps.

speed The operator usually initiates the boration mode of operation. There is no automatic actuation of the system except in the case of a volume control tank low-low level signal. In this event, the suction of the charging pump{s) is aligned to take suction from the refueling water storage ' tank which contains boron at 2500 ppm. The maximum rate of boration of the primary system with the 60 gpm discharge of a boric acid transfer pump directed to the charging pump suction is 30 ppm/minute., This provides compensation for a cooldown rate of approximately 8.8°F/minute at the end of core life when the moderator temperature coefficient is most negative. The maximum rate of boration with two of three charging pumps delivering water from the refueling water storage tank at a concentration of 2500 ppm boron is 6.2 ppm/minute. This compensates for a cooldown rate of 1.7 °F/minute at the end of core life. Alarm Functions The reactor makeup control is provided with aiat!rn functions to cal_l the operator's attention to the following conditions:

1. Deviation of reactor primary water makeup flow rate from control set point.

9 .1.2-30 12-1-69 2. 3. Deviation of concentrated boric acid flow rate from control set point. High and low level in the volume control tank. The high level alarm indicates that the ,level in the tank is approaching high level and a resulting 100% diversion of letdown stream to the Boron Recovery System. The low level alarm indicates that the level in the volume control tank is approaching low-low or emergency level in a case where the primary

    'makeup control selector is not set for the automatic makeup mode and the volume control tank level drops below the makeup initiation point.

4. Low-low level in the volume control tank.

9.1.2.4 Charging Flow Control Three single speed horizontal centrifugal charging pumps are used to supply charging flow to the Reactor *coolant System and perform the safety injection function as discussed in Sections 6.1 and 6.2. The charging mode and the safety injection mode represent two separate operating paints on the.pump head curves. A flow transmitter on the charging line upstream of the re~enerative heat exchanger transmits a signal to an indicator-controller in the control room. The controller regulates a throttling valve in the charging line to maintain a preset charging flow. A Reactor Coolant System pressurizer water level error signal resets the charging flow set point to provide corrective action. The controller is provided with adjustable maximum and minimum flow limits. Maximum flow is limited to preven~ entry into the safety injection mode and startup of the standby charging pump during normal unit transient conditions.

Minimum flow is limited to prevent flashing downstream from the letdown orifices.

9 .1. 2-31 12-1-69

The response of the charging line throttling valve to changes in the flow control signal normally is damped to reduce charging flow fluctuations due to short term pressurizer level transients. If the pressurizer level increases, the error signal changes the charging flow set point to a lower value, which causes the control valve to move towards the closed position.

A pressure switch in the charging pump discharge header actuates an alarm and starts a standby charging pump if the discharge header pressure falls to a preset low level. The safety injection signal overrides any other associated control signal. 9 .1.2 .5 Components

A summary of principal component data is given in Table 9,1-3.

Regenerative Heat Exchanger The regenerative hea 4 exchanger is designed to recover the heat from the letdown stream by reheating the charging stream during normal operation. This exchanger also limits the temperature rise which occurs at the letdown orifices during transient periods when letdown flow exceeds charging flow. The letdown stream flows through the shell of the regenerative heat exchanger and the charging stream.flows through the tubes. The exchanger is fabricated of austenitic stainless steel, and is of all welded construction. The exchanger is designed to withstand a minimum of 2000 step changes in shell side fluid temperature from 130°F to 550°F during the design life of the unit.

9.1.2-32 12-1-69 Letdown Orifices One of the three parallel letdown orifices controls flow of the letdown stream during normal operation and reduces the coolant pressure to a value compatible with the non-regenerative heat exchanger design. A second orifice is used to attain maximum purification flow at normal Reactor Coolant System operating pressure; and the third orifice serves as a spare. The orifices are placed in service by remote manual operation of their respective isolation valves. One or both of the standby orifices are used* in parallel with the normally operating orifice in order to increase letdown flow when the Reactor Coolant System pressure is below normal. This arrangement provides a full standby capacity for control of letdown flow. Each orifice is constructed from austenitic pipe containing a corrosion and. erosion resistant insert bored to the diameter required.

t~on-Regenerative Heat Exchanger The non-regenerative heat exchanger cools the letdown stream to the operating temperature of the mixed bed demineralizers. Reactor coolant flows through the tube side of the exchanger while component cooling water flows through the shell. The letdown stream outlet temperature is automatically controlled by a temperature control valve in the component cooling water outlet stream.

The unit is a multiple-pass-tube heat exchanger. All surfaces in contact with the reactor coolant are austenitic stainless steel, and the shell is carbon steel.

9. J

2:-3.3:

12-1-69 Mixed Bed Demineralizers Two flushable mixed bed demineralizers maintain reactor coolant purity by the 7 use of a Li cation resin and a hydroxyl form anion resin. These resins remove fission and corrosion products and, in addition, the borated reactor coolant converts the anio~ resin to the borate form. The resin bed is designed to reduce the concentration of ionic isotopes in the purification I stream, except for cesium, tritium and molybdenum, by a minimum factor of 10, *. Each demineralizer is sized to accommodate the maximum letdown flow. One demineralizer serves as a standby unit for use when the operating demineralizer becomes exhausted during operation *

The demineralizer vessels are fabricated of austenitic stainless steel and are provided with suitable connections to facilitate resin replacement.

The vessels are equipped with a resin retention screen. Each demineralizer* has sufficient capacity to operate for bne core cycle with one percent defective fuel rods. Deborating Demineralizers When required, two anion demineralizers remove boric acid from the Reactor Coolant System fluid. The demineralizers are intended for use near the end of a core cycle when boron concentrations ar.e low, but can. be used at any I time i f re.piired *. Hydroxyl based ion-exchange resin is used to reduce Reactor Coolant Systetr boron concentration by releasing a hydroxyl ion when borate ion is adsorbed.

9.1.2-34 12-1....,69 When the resin is saturated, it is flushed to the spent resin storage tank and new resin is added.

Each demineralizer is sized to remove that quantity of boric acid from the Reactor Coolant System necessary to maintain full power operation near the end of core life without the use of the Boron Recovery System~

Cation Bed Demineralizer A demineralizer using a flushable cation resin bed in the hydrogen form is located downstream of the mixed bed demineralizers and is used when 7 required to control the concentration of Li which builds up in the coolant from the BlO (n, a.) Li 7 reaction. The demineralizer also has sufficient capacity to maintain the cesium-137 concentration in the coolant below 1.0 µCi/cc with 1% defective fuel. The demineralizer is used to control cesium as necessary during operation. The demineralizer vessel is fabricated of austenitic stainless steel and is provided with suitable connections to facilitate resin replacement when required. The vessel is equipped with a resin retention screen. Res:i.n Fill Tank The resin fill tank is used to charge fresh resin to the demineralizers. The linei from the conical bottom of the tank is fitted with a dump valve and a flexible hose spool piece that may be connected to any one of

9.1.2-35 12-1-69

the demineralizer fill lines. The demineralizer water and resin slurry is sluiced into the demineralizer by opening the dump valve. The tank is fabricated of austenitic stainless steel and is designed to hold approximately two-thirds the resin volume of one mixed bed demineralizer.

Reactor Coolant Filter The filter collects resin fines and particulates larger than 25 microns if such fines should carry over into the letdown stream. The vessel is fab-ricated of austenitic stainless steel, and is provided with connections for draining and venting. Design flow capacity of the filter is equal to the maximum purification flow rate *

Disposable synthetic filter elements are used. Bases being considered to.

determine when the reactor coolant filter will be replaced are a high pressur.e differential across the filter, a set time limit after which the filter will be replaced and when a portable radiation monitor shows radiation in excess of established limits. Volume Control Tank The volume control tank is the collecting point in the system for letdown flow, makeup and chemical additions. It has surge capacity to compensate for changes in reactor coolant volume resulting from power level increases and the deadband in the reactor control temperature instrumentation. Over-pressure of hydrogen gas is maintained in the volume control tank to contro*l

the ijydrogen concentration in the reactor coolant at 25 - 35cc per Kg of water at standard ;temperature and pressure.

9 .1. 2-36 12-1-69 A spray nozzle is located inside the tank on the inlet line from the reactor coolant filter. This spray nozzle provides intimate contact to equilibrate the gas and liquid phases. A remotely operated vent valve discharging to the Vent and Drain __System permits_ rem~".'_al of gaseous fission products, when_~~' which are stripped from the reactor coolant at this location. The volume control tank also acts as a head tank for the charging pump suction header. The tank is constructed of austenitic stainless steel. Charging Pumps Three charging pumps inject coolant into the Reactor Coolant System. These pumps also perform the safety injection function as discussed in Sections 6.1 and 6.2. The pumps are of the single speed horizontal centrifugal type, and all parts in contact with the reactor coolant are constructed of austenitic stainless steel or other material of adequate corrosion resistance. These pumps have a mechanical seal and auxiliary gland bushing, with a leakoff connection on the auxiliary gland between the mechanical seal and bushing. This arrangement minimizes the possibility of reactor coolant leakage to the outside atmosphere. Pump leakage is piped to the primary drain header for disposal. The pump design prevents lubricating oil from contaminating the charging flow. Each pump is designed to provide the full charging flow and the reactor coolant pump seal water supply during normal seal leakage. Each pump is designed to provide rated flow against a pressure equal to the sum of the Reactor Coolant System normal maximum pressure and the piping, valve and equipment

9 .1. 2-37

                                                                   *i2::-1-69 pressure losses at the design charging flows. The capacity of each charging

pump permits operating at normal charging line flow with one reactor coolant pump shaft seal operating normally while the other two reactor coolant pumps are operating with significant floating ring seal flow. The capacity of each pump also provides sufficient bypass recirculation to prot~ct the pumps by preventing blocked discharge during testing and too low a flow at minimum charging conditions.

Chemical Mixing Tank The primary use of the chemical mixing tank is for the preparation of solutions for pH control and oxygen scavenging and it has a capacity more than sufficient to pr.epare a solution of pH control chemical for the Reactor Coolant System. It is fabricated of austenitic stainless steel. The capacity of the chemical mixing tank is determined by the quantity of 35% hydrazine solution necessary to increase the concentration in the reactor coolant by 10 ppm. Excess Letdown Heat Exchanger The excess letdown heat exchanger cools reactor coolant letdown flow if letdown through the normal letdown path is blocked. It is designed to cool a letdown flow equal to the nominal injection rate through three reactor coolant pump l~byrinth seals. The unit is designed to reduce the letdown stream temperature from the cold leg temperature to 195°F. The letdown stream flows through the tube side and component cooling water circulates through the shell side. All surfaces in contact with the reactor coolant are austenitic

stainless steel, and the shell is carbon steel. All tube joints are welded
The unit is designed to withstand 12,000 step changes in the tube fluid temperature from 80°F to the cold leg temperature.

9 .1. 2-38 12-1-69 Seal Water Heat Exchanger

The seal water heat exchanger removes heat from three sources: reactor coolant pump seal water~reactor coolant discharged from the excess letdown heat exchanger and charging pump :bypas~-- fl~. Reactor coolant flows th:i;ougl:!_ ______ _

the tubes and component cooling water is circulated through the shell side. The tubes are welded to the tube sheet because undesirable leakage could occur in either direction. All surfaces in contact with reactor coolant are austenitic stainless steel, and the shell is carbon steel. The exchanger is designed to cool the excess letdown flow and the seal water flow to the temperature normally maintained in the volume control tank if all the reactor coolant pumps floating ring seals are leaking at the maximum design leakage rate. Seal Water Filter The filter collects particulat~s larger than 25 microns from the reactor coolant pump seal water return and from the excess letdown heat exchanger flow. The filter is designed to pass the sum of° the excess letdown flow and the maximum desi~ leakage from the reactor coolant pump floating ring seals. The vessel is constructed of austenitic stainless steel and is provided with connections for draining and venting. Disposable synthetic filter elements are used.

9 *.1.2,-3"9 12-1-"69 Boric Acid Filter

The boric acid filter collects particulates larger than 25 microns from the boric acid solution being pumped to *the charging pump suction ~ine or boric acid blender. The filter is designed to pass the design flow of two boric acid pumps operating simultaneously. The vessel is construc.ted of austenitic stainless steel and the filter elements are disposable synthetic cartridges. Provisions are available for venting and draining the filter.

Boric Acid .Tanks Boric acid solution recovered from the Boron Recovery System or mixed in the batching -.tank is stored in three electrically heated boric acid tanks shared by both units. One tank for each unit supplies boric acid for reactor coolant makeup and for circulation through that units boron injection tank while recycled solution from the boron recovery evaporator is being accumulated in the third tank. The three tanks combined have sufficient boric acid capacity to provide cold shutdown for the two units, each with one control rod assembly completely withdrawn, following a refueling shutdown on both uni.ts. Each tank, .if maintained above the low level alarm point, can supply sufficient boration to provide cold shutdown for one unit with a control rod assembly completely withdrawn. The concentration of boric acid solution in storage is maintained between 11.5 and 12.5% by weight. Periodic manual sampling and corrective action, if necessary, ensure that these limits are maintained. As a consequence, measured amounts of boric acid solution could be delivered to the reactor coolant to control the boron concentration. Each boric acid tank has an overflow with a water loop seal which is connected to the Waste Disposal System. The tanks are constructed of austenitic stainless steel.

9 .1. 2-40 12-1-69 Batching Tank The batching tank is sized to hold one week's makeup supply of boric acid solution for transfer to the boric acid tanks. The basis for makeup is reactor coolant leakage of 1/2 gpm at beginning of core iife. The tank may also be used for solution storage . . A local sampling point is provided for verifying the solution concentration prior to transferring it. to the. boric acid tank or for draining the tank. A tank manway is provided with a removable screen to prevent entry of foreign particles. In addition, the tank is provided with an agitator to improve mixing during batching operations. The tank is constructed of austenitic stainless steel and is not used to handle radioactive substances. The tank is provided with a steam jacket for heating the boric acid solution to 165°F. Boric Acid Tank Heaters Two 100% capacity electric immersion heaters in each boric acid tank are designed. to maintain the temperature of the boric acid solution at 165°F with ambient air temperature of 40°F thus ensuring a temperature in excess of the solubility limit (for 21,000 ppm boron this is 135°F). The heaters are sheathed in austenitic stainless steel. Boric Acid Transfer Pumps Four centrifugal two sp~ed pumps are used to circulate' or transfer the boric acid solution. The pumps circulate boric acid sdiution. through the boric .acid

9.1.2-41 12-1

tanks and the boron injection tank in the Safety Injection System, and inject boric acid: into the charging pump suction header or furnish boric acid to the boric acid blender. Although one pump normally is used for boric acid batching and transfer for each unit and one for boric acid injection for each unit, either pump may function as standby for the other. The design capacity of each* pump is equal to the normal letdown flow rate. The d~sign -

head of one pump is sufficient, considering line and valve losses, to deliver rated flow to the charging pump suction header when volume control tank pressure is at the maximum operating value. All parts in contact with the solutions are austenitic stainless steel or other suitable corrosion-resistant material. The boric acid transfer pumps are operated either automatically or manually from the main control room. The reactor makeup control operates one of the pumps automatically when boric acid solution is required for makeup or boration. Boric Acid Blender The boric acid blender promotes thorough mixing of concentrated boric acid' solution and primary grade water for the reactor coolant makeup circuit. The blender consists of a conventional pipe fitted with a perforated tube insert. All material is austenitic staipless steel. The blender decreases the pipe length required to homogenize the mixture for taking a representative* local sample .

9.1.2-42 12-1-69 Electrical Heat Tracing Electrical heat tracing is installed under the insulation on all pumps, piping, i valves, line-mounted instrumentation, and components normally containing concentrated boric acid solution. The heat tracing is designed to prevent boric acid precipitation due to cooling, by compensating for heat loss. Exceptions are:

1. Lines which may transport concentrated boric acid but are subsequently flushed with reactor coolant or other liquid of low boric acid concentration during normal operation.

2. The boric acid tanks which are provided with immersion heaters.

3.. The batching *tank, which is provided with a steam jacket. Duplicate tracing on sections of the Chemical and Volume ~ontrol System normally containing boric acid soluti?n provides backup if the operating tracing malfunctions. The existence of a condition which requires a redundant tracing to be operated will be indicated by an alarm in the Main Control Room. Valves Valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection that discharges directly, or via a floor drain, to the Vent and Drain System. All other valves have stem leakage control. Globe valves are installed*

9.1.2-43 12-1-69 * -*-. with flow over the seat when such an arrangement reduces the possibility of leakage. Basic material of construction is stainless steel for all valves except the batching tank steam jacket valves which are carbon steel. Isolation valves are provided at all connections to the Reactor Coolant System. Connections to the Reactor Coolant System which pass through.the containment are equipped with isolation devices as described in Section 5.2. Relief valves are provided for lines and components that might be pressurized above design pressure by improper operation or component malfunction. Pressure relief for the tube side of the regenerative heat exchanger is provided by a spring loaded check valve around the charging line isolation valve. The valve relieves to the Reactor Coolant System. All relief valves used in systems handling radioactive fluids are of the closed bonnet design and are constructed of stainless steeL Piping All Chemical and Volume Control System piping handling radioactive liquid is austenitic stainless steel. All piping joints and connections are welded, except where flanged connections are required to facilitate equipment removal for maintenance and hydrostatic testing. Piping, valves, equipment and line-mounted instrumentation, which normally contain concentrated boric acid solution, are heated by electrical tracing to ensure solubility of the boric acid .

9.1.3-,.l 12-1-69 SYSTEM DESIGN EVALUATION 9.1. 3 9.1.3.1 Availability and Reliability A high degree of functional reliability is assured in this system by providing standby components where performance is vital to safety and by assuring fail-safe response to the most probable mode of failure. Special provisions include duplicate heat tracing with alarm protection of lines, valves and components normally containing concentrated boric acid. The Chemical and Voltnne Control System has three high pressure charging pumps, each capable of supplying the required reactor coolant pump seal and makeup flow

lhe electrical equipment of the Chemical and Volume Control System is arranged so that multiple items receive their power from two 480 volt buses (see Figure 8.1-1). Two of the charging pumps and one of the boric acid transfer pumps are powered from a separate 480 volt bus than the remaining charging pump and boric acid transfer pump. In case of loss pf a-c power, a charging pump and a boric acid transfer pump can be placed on the.emergency di~sels,if necessary.

9,1.3.2 Control of Tritium The Chemical and Volume Control System is used to control the concentration of tri.tium in the Reactor Coolant System. Essentially all of the

9.1.3-2 12~1-69 tritium is in chemical combination with oxygen as a form of water. Therefore, any leakage of coolant to the containment atmosphere carries tritium in the same proportion as it exists in the coolant. Thus, the level of tritium in the containment atmosphere, when it is sealed from outside air ventilation, is mainly a function of tritium level in the reactor coolant. In addition it depends on the cooling water temperature at the ventilation cooling coils, and the presence of leakage other than reactor coolant as a source of moisture in the containment air *. There are two major considerations with regard to the presence of tritium in the reactor coolant:

1. Possible station ~ersonnel hazard during access to the containment.

Leakage of reactor coolant during operation causes an accumulation of tritium in the containment atmosphere. 2 *. Release of tritium to the environment. I Neither of these considerations are limiting in the operation of the unit. 9.1.3.3 Leakage Provisions All Cfteinical and Volume Control System valves and piping for radioactive services are designed to permit essentially zero leakage. The components designated for radioactive service are provided with welded connections

9.1.3-'-3' 12-1-69 to prevent leakage. However, flanged connections are provided on

each charging pump suction and.discharge, on Jach boric acid pump suction and discharge., on the relief valve inlets and outlets,* on three-way valves and on the flow meters to permit removal for maintenance.

The centri"fugal charging pumps are provided with leakoffs to control reactor coolant*leakage, if any. All valves which are larger than 2 inches and which are designated for radioactive service at an operating fluid temperature above 212°F are provided with a stuffing box and lantern le~koff connections. All control valves are provided with. stuffing box and leakoff connections or are totally enclosed, and leakage is essentially* zero for these valves. Diaphragm valves are provided where the operating pressure is 200-

 *psig or below and operating temperature is    200°F. or below. Leakage is essentially zero for these. valves.

9.1.3.4 Incident Control The letdown line and the reactor coolant pump seal water return lines penetrate* the reactor containment. The letdown line contains air-operated*valves inside the reactor containment and one air-operated valve outside the reactor containment which is automatically closed by the* containment isolation signal. The reactor coolant pump seal water return lines contain one motor-

operated isolation valve outside the reactor containment which is automatically closed by the containment isolation signal.

9 .1. 3-4 12-1-69 The seal water injection lines to the reactor coolant pumps and the charging line are inflow lines penetrating the reactor containment. Each line contains two check valves in series inside the reactor .containment to provide isolation of the reactor containment should a break occur in these lines outside the reactor containment. 9.1.3.5 Malfunction Analysis. To evaluate system safety, failures or malfunctions are assumed concurrent with. a loss--of-coolant accident, and the consequences are analyzed.

Proper* cons:idera.tion is given ta station safety in the design of the system. Result~ of this analysis are presented in Table 9.1-8.

If a rupture takes place between a reactor coolant loop and the first isolation valve or check valve, a loss of reactor coolant occurs. The first isolation or check valve is always located as.close as possible to the reactor coolant *1oop pipe. The analysis of loss-of-coolant accident is discussed in Section 14. If a rupture occurs in the Chemical and Volume Control System outside the containment, or at any point beyond the first check valve or remotely operated isolation valve 7 actuation of the valve limits the release of coolant and assures continued functioning of the normal means of heat dissipation fromithe* core. For the general case of rupture i

9 .1. 3-5 12-1-69 TABLE 9.1-8 CONSEQUENCE,:_~ *oF

FAILURES_ O~MALFUNCTI9!!S OF THE CHEMICAL AND_ VOLUME CONTROL SYSTEM WITHIN THE REACTOR CONTAINMENT Component Failure Comments and Consequences

a. Letdown.Line Rupture in the The remote air-operated valve line inside located near the main coolant loop the reactor .is closed on low pressurizer level containment to prevent supplementary loss o*f coolant through the letdown line rupture. The containment isolation valve in the letdown line outside the reactor containment is automatically closed by the containment isolation signal initiated by the concurrent loss-of-coolant accident. The closure of that valve prevents any leakage of the reactor containment atmos.phere outside the reactor containment.

b. *Charging Line Rupture in the The check valve located near the line .inside main coolant loop prevents the reactor supplementary loss of coolant

containment through the line rupture.

The air-qperated valve

Component TABLE 9 .1-8 ( Continued) .. Failure Comments and Consequences located upstream of the. check. valve in the defective line closes to isolate the Reactor Coolant System from the rupture. The check valve located at the boundary of the reactor containment prevents any leakage of the reactor containment atmosphere outside the reactor containment.

c. Seal Water Return Line Rtip~ure in the _

line inside The motor-operated isolation valve located outside the containment is the reactor manually closed or is automatically containment closed by the containment isolation signal initiated by the concurrent loss-of-coolant accident. The closure of that valve prevents any leakage of the reactor containment atmosphere outside the reactor containment.

9.1.3-7 4-15-70 - outside the contaimnent, the largest source of radioactive fluid subject to release is the contents of the volume control tank. of such a release are discussed in Section 14. The consequences When the reactor is subcritical during cold or hot shutdown, refueling, and approach to criticality, the relative reactivity status is continuously monitored and indicated by Bio counters and count rate indicators. Any appreciable increase in the neutron source multiplication, including that caused by the maximum possible boron dilution rate of approximately 350 ppm per hour, is slow enough to give ample time to start corrective action to prevent the core from becoming critical. The maximum dilution rate is based on the abnormal condition of two charging pumps delivering unborated - makeup water to the Reactor Coolant System at a particular time during

refueling when the boron concentration is at the maximum value and the water volume in the system is at a minimum.

At least two separate and independent flow paths are available for normal reactor coolant boration; i.e., the charging line, or the reactor coolant pum~ seal labyrinths. The malfunction or failure of one in either component does not result in the inability to borate the Reactor Coolant System. An. alternate flow path is always available for emergency boration of the reactor coolanf. As backup to the boration system, the operator can also align the refueling water storage tank outlet to the suction of the charging pump, if required.

9.i.J-8 12-1-69 A single malfunction in one of the boron makeup subsystems does not preclude the ability to maintain proper boron concentration in both units simultaneously. Subsequent to complete loss of seal injection water to the reactor coolant I pump seals, low charging pressure in the system header (below a preset value) automatically starts a standby charging pump. Even if the seal water injection flow is not re*-established, the unit can operate* indefinitely if component cooling water is available since the thermal barrier cooler has sufficient _capacity to cool the reactor coolant flow which passes through the thermal barrier cooler and seal leakoff from the pump volute. i 9.1.3.6 Galvanic Corrosion

                                                                                  . I The only types of materials which are in contact with each other in borated water are stainless steels, Inconel, Stellite valve materials and Zircaloy fuel element cladding. These materials have been shown(l) to exhibit only an insignificant degree of galvanic corrosion when coupled to each other.

For example, the galvanic corrosion of Inconel versus 304 stainless steel resulting from high temperature tests (575°F) in lithiated, boric*, acid 2 solution was found to be less than -20.9 mg/dm for the test period of 9 days. _Further galvanic corrosion would be trivial since the cell currents at the conclusion of the tests were approaching polarization. Zircaloy versus 304 stainless steel was shown to polarize at 180°F with lithiated, boric

     .                                                      .                2 acid solution in less than 8 days with a total galvanic attack of -3.0 mg/dm.

9 .1. 3-9 12-1-69

Stellite versus 304 stainless steel was polarized in 7 days at 575°F in lithiated boric acid solution.

2 (1) was -0.97 mg/dm. The total galvanic corrosion for this couple These tests show that the effects of galvanic corrosion are insignificant in systems containing borated water. (1) WCAP 1844 "The Galvanic Behavior of Materials in Reactor Coolants" D. G. Sammarone, August, 1961 .

9 .1.4-1 12-1-69 9.1.4 MINIMUM OPERATING CONDITIONS The minimum operating conditions for the Chemical and Volume Control System are contained in the Technical Specifications

9.1.5-1 12-1-69 9.1.5 TESTS AND INSPECTIONS Periodic testing, calibration, and inspection are conducted on the various instrument channels to assure proper instrument response and operation of alarm functions. The minimum frequency for testing, calibrating and for inspection are contained in the Technical Specifications.

Most components are in use regularly during power operation; therefore, assurance of the availability and performance of the system and equipment is provided .

FIG.9.1-1

OCT. 15.1970

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8. PROVlDE' MIWIMUM OF" ~TRAIGHT ~ UN

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FIG. 9.1-2 OCT. 15,.1970

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FIG. 9.1-3 OCT. 15,1970

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FIG.9.1-4 OCT. 15, 1970 NOTES;

j. VALVE FAil WITH FLOW TO VOLUME. CONTROL TAN"'-

1, SPECIAL VALVE* FUNCTIONS AS ~OT"H ISOLATIOH AND RELIEF VALVE. . 3, ELECTROMAGNETIC LOADED METER IN VERTICAL. PIPE RUN.

4. (',LOBE VALVES ARE NORMALLY INSTALLED W 1TH FL0\11 UNDER SEAT.

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7. ECCENTRIC REOUCER. 5TRAIGHT SIPE INSTALi-ED AT TOP Of" PJ?E' RUN.

8. MECUANICAL SEAL TELL-TALE CONNCCTIONS ON I a"-CHca87-IS2. -s'~Cl-\-5SHS'Z. AUX GLAND PLATE: VE:NT C.ONNEC.TIOIIS ARE PWIGGED.

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9.2.1-1 12-1-69 9.2 BORON RECOVERY SYSTEM Th,e,...ffor;;-~~shown in Figures 9.2-1, 9.2-2, 9.2-3, and 9.2-4 is a

\_

co~on system serving both units. The system degasifies and stores borated radioactive water letdown by the Chemical and Volume Control System (Section 9.1). Further processing by evaporators, filters, and demineralizers in this system produc,es primary grade water and concentrated boric acid solution for statiop. reuse or disposal. The system is designed for liquid samples to be taken before distillate is reused or sent to the Liquid Waste Disposal System. 9.2.1 DESIGN BASIS

The Boron Recovery System capacity is sufficient to accommodate the coolant let-down flow J)roduced by two cold shutdowns from full power in one unit plus on:e

  - - - - ~ - - - - - ----- --- -- --- ----* ----------- ----~

cold shutdown from full power in the second unit in a seven day period. These shutdowns are assumed to occur at that point in core life when the operating boron concentration in the first unit is 100 ppm and the boron concentration fn the second unit is one month out of phase. The system influent results from shutdown berating bleed, draining one reactor coolant* loop for maintenance woirk,- system expansion *during heatup, and dilution bleed to operating boron concentra-- tion on startup. The boron recovery tanks are assumed to be 10 percent full at the time of a cold shutdown and the boron evaporators 75 percent available at rated capacity during the period. The Boron Recovery Sys-tern can accommodate letdown flow due to daily load fei))ow-

ing and weekend. load reductions on both units to nearly the end of core life*

9.2.1-2 12-1-69 with 7S percent evaporator availability with minimum use of boron recovery tank capacity. The daily load follow cvcle basis consists of 12 hr at full power, a uniform 3 hr ramp reduction to SO percent power, 6 hr at SO percent power and a uniform 3 hr ramp increase to full power. The system is capable of removing gases from both units simultaneously at the maximum letdown flow rate. The boron evaporators are capable of processing the average letdown rate of both units producing a distillate with boron content not exceeding 10 ppm boron and concentrated bottoms at 12 percent boric acid. The primary drain tank, gas stripper, gas stripper overhead condenser, primary drain tank vent chiller condenser, overhead gas compress~rs, and gas surge tank in the Boron Recovery System are designed as Class I components. All piping in the Boron Recovery System is type.304 stainless steel. All piping joints and connections are welded except where flan~ed connections are required to facilitate equipment removal for maintenance. All valves handling radioactive gas are packless, diaphragm valves. All valves handling primary grade water or radioactive fluid are stainless steel. All liquid lines, equipment, and accessories containing concentrated boric acid (6 percent by weight boric acid or greater) are electrically heat traced, with dual circuits to prevent crystallization of boric acid. The boron recovery tanks

9.2.1-J 12-1-'19

and primary grade water tanks are heated by steam. The evaporator bottoms tank is maintained at 150° F minimum by dual electric heaters.

The design data for the Boron Recovery System components are given in Table 9. 2-1.

9.2.1-4 12-1-69 TABLE 9.2-1 BORON RECOVERY SYSTEM COMPONENT DESIGN DATA Primary Drain Tank Number 1 Capacity, gal 5,000 Design Pressure, psig 30 Design Temperature, °F 240 Operating Pressure, psig 2 0 Operating Temperature, F 125 Material Stainless Steel Type 304 Design Code ASME III C Gas Stripper Number 1 1,248 Capacity, gal Design Pressure, psig 50 Design Temperature, °F 300 Operating Pressure, psig 2 0 Operating Temperature, F 219 Material Stainless Steel Type 316L Design Code ASME III C

9, *. 2. l;-5. 12-1:..,.69

- Boron Recovery Tanks TABLE 9 *. 2-1 (Continued)

Number 3 (one required for each u~it;) Capacity, gal 120,000 Design Pressure, psig o.s Des i gn Tempe.rature, OF 180 Operating Pressure Atmospheric Operating Temperature, °F 130 Material Stainless Steel Type 3Q4L Design Code API-650 (See Note 1) Boron Evaporators Number 2 Capacity, each, gal 2,900 Design Pressure, psig 100 Des i gn Temperature, OF 338 Operating Pressure, psig 15 0 Operating Temperature, F 260 Material Stainless Steel Type 316L Design Code ASME III C Note 1 - In addition to the API-650 Code, the construction incorporated requirements of ASME Code Section III C, for welding, welding procedure qualifications, weld joint efficiency and weld inspection. '

9.2.1-6 12-1-69 Primary Water Tanks TABLE 9.2-1 (Continued) Number 2 Capacity, each, gal 180,000 Design Pressure, psig 0.5 Des i gn Temperature, OF 140 Operating Pressure Atmospheric 0 perat i ng T emperature, OF 125 Material Stainless Steel Type 304L Design Code API-650 (See Note 1) Evaporator Bottoms Tank Number Capacity, gal Design Pressure, psig 1 4,000 25 Des i gn T emperature, OF 200 Operating Pressure Atmospheric Operating Temperature, °F 160 Material Stainless Steel Type 316L Design Code ASME III C

1 9.2.1-7 12-1-69 TABLE 9.2-1 (Continued) Distillate Accumulators Number 2 Capacity, each, gal 550 Design Pressure, psig 100 Design Temperature, °F 338 Operating Pressure, psig 15 0 Operating Temperature, F 250 Material Stainless Steel Type 304 Design Code ASME III C

Gas Surge Tank Number Capacity, gal 1

525 Design Pressure, psig 200 Design Temperature, °F 200 Operating Pressure, psig 120 0 Operating Temperature, F 150 Material Stainless Steel Type 304 Design Code ASME III C

9.2.1-8 12-1-69 TABLE 9.2-1 (Continued) Test Tanks Number 2 Capacity, each, gal 30,000 Design Pressure, psig 0.5 Des i gn Temperature, OF 140 Operating Pressure Atmospheric Operating Temperature, °F 125 Material Stainless Steel Type 304L Design Code API-650 (See Note 1) Stripper Feed Heat Exchangers . Number Total Duty, Btu/hr 2 18,000,000 Shell Tube Design Pressure, psig 200 150 Design Temperature, OF 300 200 Operating Pressure, psig 125 100 Operating Temperature, In/Out, OF 219/144 100/175 Material ss 304 ss 304 Fluid Letdown Letdown Design Code ASME III C ASME III C

9.2.1-9

                                                                    .12-1-69

TABLE 9.2-1 (Continued)

Stripper Feed Steam Heaters Number 2 Total Duty, Btu/hr 7,800,000 Shell Tube Design Pressure, psig 200 150 Design Temperature, °F 388 338 Operating Pressure, psig 100 100 Operating Temperature, In/Out, °F 338/338 175/240 Material Carbon Steel ss 304 Fluid Steam Letdown

Design Code Stripper Trim Cdoler ASME VIII ASME III C Number 1 Total Duty, Btu/hr 1,700,000 Shell Tube Design Pressure, psig 150 200 Design Temperature, °F 150 220 Operating Pressure, psig 75 75 0

Operating Temperature, In/Out, F 105/112 144/ 130 Material Carbon Steel ss 304 Fluid Component. Letdown Cooling Water

Design Code ASME III C ASME III C

9. 2 .1-'10 12-1-69 TABLE 9.2-1 (Continued)

Stripper Overhead Condenser Number 1 Total, Btu/hr 2,800,000 Shell Tube Design Pressure, psig 150 150 Design Temperature, °F 300 300 Operating Pressure, psig 2 75 0 Operating Temperature, In/Out, F 219/219 105/116 Material ss 304 ss 304 Fluid Distillate Component Cooling Water Design Code Primary Drain Tank Vent Chiller Condenser ASME III C ASME III C Number 1 Total Duty, Btu/hr 20,000 Shell Tube Design Pressure, psig 150 150 Design Temperature, °F 300 300 Operating Pressure, psig 2 75 Operating Temperature, In/Out, °F 219/130 75/77 Material ss 304 ss 304 Fluid Distillate Chilled Component Cooling Water Design Code ASME III C ASME III C

9.2.1-11** 10-15;_70

Boron Evaporator Reboilers TABLE 9.2-1 (Continued)

N\llllber 2 Duty, each, Btu/hr ll, 100,000 Shell Ttibe Design Pressure, psig 200 100 Des i gn Temperature, OF 382 300 Operating Pressure, psig 100 25 Operating Temperature, In/Out, °F 338/338 253/263 Material Carbon Steel ss 304 Fluid Steam 1 - 12% Boric Acid Design Code ASME VIII ASME III C Boron Evaporator Distillate Coolers Number 2 Duty, each , Btu/hr 1,150,000 Shell Tube Design Pressure, psig 100 150 Design Temperature, °F 338 338 Operating Pressure, psig 50 75 Operating Temperature, In/Out, °F 240/125 105/ 139 Material 304 ss 304 ss Fluid Distillate Component Cooling Water Design Code ASME III C ASME III C

9.2.1-12 12-1-69' TABLE 9.2-1 (Continued) Boron Evaporator Bottoms Cooler Number 1 Total Duty, Btu/hr ,950,000 Shell Tube Design Pressure, psig 150 150 Design Temperature, ~F 300 300 Operating Pressure, psig 85 45

               .          .. 0 Operating Temperature, In/Out,' F               150/170        150/160 Material                                       Carbon Steel    ss  304 Fluid                                          Component       12% Boric Acid Cooling Water Design Code Boron Recovery Tank Heaters ASME III C      ASME III C Number                                         3 Duty, Each, Btu/hr                             670,000 Shell           Tube Design Pressure, psig                          200             200 Des i gn Temperature, OF                       388             388 Operating Pressure, psig                        100            30 Operating Temperature, In/Out, °F              338/338         40/250 Material                                       Carbon Steel    ss  304 Fluid                                          Steam           Letdown Design Code                                    ASME VIII       ASME III C

9. 2,.1-13 12-1-69

Primary Water Tank Heaters TABLE 9.2-1 (Continued)

Number 2 Duty, each, Btu/hr 635,000 Shell Tube Design Pressure, psig 200 200 Des i gn Temperature, OF 388 388 Operating Pressure, psig 100 31 Operating Temperature, In/Out, °F 338/338 40/252 Material Carbon Steel ss 304 Fluid Steam Water

Design Code Primary Drain Tank Pumps ASME VIII ASME VIII Number 2 (one required)

Type Horizontal centrifugal Motor Horsepower, hp 20 Seal Type Canned Pump Capacity, each, gpm 240 Head at Rated Capacity, ft 222 Design Pressure, psig 150 Materials Pump Casing ss 316 Shaft ss 316

- Impeller ss 316

9.2.1-14 12-1-69 TABLE 9.2-1 (Continued) Gas Stripper Circulating Pumps Number 2 (one required) Type Horizontal Centrifugal Motor Horsepower, hp 30 Seal Type Mechanical with Backup Breakdown* Section Capa~i ty, each., gpm 240 Head at Rated Capacity, ft 257 Design Pressure, psig 225 Material Pump Casing ss 316 Shaft SAE 4140 Impeller ss J16 Boron Evaporator Feed Pumps Number 2 (one required) Type. Horizontal Centrifugal Motor Horsepower, hp 10 Seal Type Mechanical Seal with Backup Breakdown Section Capacity, each, gpm 150 Head at Rated Capacity, ft 117 Design Pressure, psig 225 Materials Pump Ca~ing ss 316 Shaft SAE 4140 Impeller ss 316

9.2.1-15 12-1-69

TABLE 9.2-1 (Continued)

Boron Evaporator Circulating Pumps Number 2 Type Horizontal Centrifugal Motor Horsepower, hp 20 Seal Type Double Mechanical Capacity, each, gpm 2,200 Head at Rated Capacity, ft 20 Design Pres~ure, psig 225 Materials Pump Casing ss 316 Shaft SAE 4140 Impeller ss 316 Boron Evaporator Bottoms Pump~ Number 2 (one required) Type Horizontal Centrifugal Motor Horsepower, hp 1 1/2 Seal Type Canned Pump Capacity , each., gpm 20 Head at Rated Capa.city, ft 56 Design Pressure, pstg 150 Materials Pump Casing ss 316 Shaft ss 316

Impeller ss. 316

9.2.1-16 12-1-69 TABLE 9.2-1 (Continued) Boron Evaporator Bottoms Cooler Circulating Pump Number 1 Type Horizontal Centrifugal Motor Horsepower, hp 1 1/2 Seal Type Mechanical Capacity, each, gpm 50 Head at Rated Capacity, ft 30 Design Pressure, psig 150 Materials Pump Casing Cast Iron Shaft Carbon Steel Impeller Cast Iron Boron Evaporator Bottoms Tank Circulating Pump Number 1 T:rpe Horizontal Centrifugal Motor Horsepower, hp 1 1/2 Seal Type Canned Pump Capacity, each, gpm 50 Head at Rated Capacity, ft 52 Design Pressure, psig 150 Materials Pump Casing ss 316 Shaft ss 316 Impeller ss 316

9.2~ 1:-17 12-1-6,9

TABLE 9.2-1 (Continued)

Boron Evaporator Distillate Pumps Number 2 Type Horizontal Centrifugal Motor Horsepower, hp 5 Seal Type Mechanical Capacity, each , gpm 22 Head at Rated Capacity, ft 140 Design Pressure, psig 225 Materials Pump Casing ss 316 Shaft SAE 4140

Impeller Test Tanks Pumps ss 316 Number 2 (one required)

Type Horizontal Centrifugal Motor Horsepower, hp 10 Seal Type Mechanical Capacity, each; gpm 100 Head at Rated Capacity, ft 142 Design* Pressure, psig 225 Materials Pump Casing ss 316 Shaft SAE 4140 Impeller ss 316

9.2.1-18 12-1-69 Primary Water Supply Pumps TABLE 9. 2-1 (Continuedi Number 2 (one required) Type Horizontal Centrifugal Motor Horsepower, hp 30 Seal Type Mechanical Capacity, each, gpm 350 Head at Rated Capacity, ft 255 Design Pressure, psig 225 Materials Pump Casing ss 316 Sha~t SAE 4140 Impeller ss 316 Overhead Gas Compressor Number 2 (one required) Type Diaphram Motor Horsepower, hp 2 Capacity, each, scfm 2.5 Discharge Pressure at Capacity, psig 125 Design Pressure, psig 200 Materials Cylinder Carbon Steel Piston Rod Forged Steel Piston Nodular Iron Diaphram and Parts Contacting Gas 302/304 or 316 SS

9.2.1--19 12-1-69

Boron Recovery Filters Number TABLE 9.2-1 (Continued) 2 (one required)

Retention Size, Microns 1-3 Filter Element Material Fibre Capacity Normal, gpm 240 Capacity Maximum, fpm 300 Housing Material ss 304 Design Pressure, psig 150 Design Temperature, °F 250 Design Code ASME III C

Boron Evaporator Bottoms Filters Number Retention Size, Microns 2 (one required) 25 Filter Element Material Fibre Capacity Normal, gpm 20 Capaci.ty Maximum, gpm 50 Housing Material ss 304 Design Pressure, psig 150 Des i gn T emperature, OF 250 Design Code ASME *rn C

9.2.1-20 12-1-69 TABLE 9.2-1 (Continued) Boron Cleanup .Filter Number 1 Retention Size, Microns 5 Filter Element Material Fibre Capacity Normal, gpm 100 Capacity Maximum, gpm 130 Housing Material ss 304 Design Pressure, psig 150 Des i gn Temperature, OF 250\ Design Code ASME-III C Cesium Removal Ion Exchangers Number Design Flow; gpm/sq ft 2 (one required) 25 Resin Type Cation, Mono Bed 3 45 Resin Active Volume, ft Design Pressure, psig 200 Design Temperature, °F 250 Material ss 316 Design Code l\SME III C

9.2.1-21 12-1-69

Boron Cleanup Ion Exchanger TABLE 9.2-1 (Continued)

Number 2 (one required) Design Flow, gpm/sq ft 10. 5 Resin Type Cation-Anion, Mixed Bed 3 Resin Active Volume, ft 45 Design Pressure, psig 200 Design Temperature, °F 250 Material ss 316 Design Code ASME III C

Note 1 - In addition to the API-650 Code the construction incorporated requirements of ASME Code Section III C, for welding, welding procedure qualifications, weld joint efficiency and weld inspection
9.2.2-1 12-1-69
9.

2.2 DESCRIPTION

The.Boron Recovery System is illustrated on Figures 9.2-1, 9.2-2, 9.2-3, 9.2-4. Reactor coolant letdown, with entrained hydrogen and fission gases, enters the Boron Recovery System via the Vent and Drain System (Section 9.7). This liquid is pumped under automatic level control from the primary drain tank to the gas stripper, stripped of dissolved gases, .and, if necessary, passed through ion exchangers for the removal of soluble fission and corrosion products. After subsequent filtration to remove additional particulate materials, the liquid is held up in the three boron recovery tanks .for processing in the boron recovery evaporators., Noncondensable gases removed in the gas stripper are taken off the gas stripper overhead condenser and discharged into the gas stripper surge tank

. by the overhead gas compressors. The surge tank discharges either to the volume control ,tank to return the hydrogen and radioactive gases to the Reactor Coolant System, (Section 4.0), or to the Gaseous Waste Disposal System discussed in Section 11.2.5. The surge tank contains sufficient gas to provide a cover gas for the gas stripper to prevent drawing in air which could form a combustible mixture when the stripper is shut down.

The boron evaporators are fed from the boron recovery tanks by the boron evaporator feed pumps. The evaporators are of the external reboiler type and have a separation factor 5 of about 10 4 to 10

Steam from.the evaporators flows to the boron evaporator overhead condensers. After being condensed, the condensate .is held in distilla1te

9.2.2-2 12-1-69 accumulators and then pumped through the distillate coolers to the test tanks for sampling. The distillate accumulator vents discharge to the Gaseous Waste Disposal System discussed in Section 11.2.5 only when noncondensable gases build up in the distillate accumulator. Liquid from the test tanks is pumped to the primary grade water storage tanks, passing through a filter and an ion exchanger for further boron removal, if required. If unacceptable or not required for reuse, the test tank liquid is pumped to the Liquid Waste Disposal System (Section 11.2). The primary grade water requirements of both units are supplied from two tanks using associated pumps. When the concentration of boric acid in the bottom of the boron evaporator is at ~ the recovery concentration of approximately 12 percent, the liquid is pumped through a controlled temperature bottoms cooler and a filter to the boron evapo-rator bottoms tank. The concentrate in the tank is sampled and if disposal of bottoms is desired, the liquid is transferred to the drumming station in the Waste Disposal System as described in Section 11.2.4. The Boron Recovery System is designed so that operation of the primary drain tank and gas stripper is automatic when all system control set points are established. Operation of the evaporators is automatic upon cycle initiation from the Main Control Room.

9 *.2.3-1 12-1-69

9.2.3 DESIGN EVALUATION The design capacity of the gas stripper is 240 gpm, which corresponds to the maximum instantaneous letdown rate .of both units. The strip.per is controlled automatically at .any letdown rate up to its maximum with no operator.action.

The boron recovery tanks, when 10 *percent full, have an additional capacity .of 324,000 gal. During .seven days of operation at 75 percent availability, the evaporators can process 450,000 gal of water. This provides a total capability of 624,000 gal of letdown which can *be stored or processed during any seven day period. This capability is in excess of the*450,000 gal which are produced l>Y 3 cold shutdowns *

9.2.3.1 System .Reliability Duplicate, .full capacity pumps and compressors are provided for all equipment except the boron evaporator recirculation pumps all:d bottoms tank recirculation pump. The primary grade water pumps, primary drain tank pumps, gas stripper pumps, and .gas stripper overhead compressor are provided with automatic contr,ol:s to start the standby pump if the normal pump fails. The controls of all dup:J:.:J,;..

cate pumps are designed to permit alternate duty to equalize operating hours. This system is designed as Class I to resist earthquakes and is protected from possible tornado missiles by concrete walls or ceilings

9.2.3-2 12-1-69 9.2.3.2 Malfunction Analysis A failure analysis of Boron Recovery System components is presented in Table 9.2.3-1.

9.2.3-3 12-1-69

TABLE 9.2.3-1 BORON RECOVERY SYSTEM MALFUNCTION ANALYSIS Component Malfunction Comments and Consequences Tanks and other Leak Tanks and other components components con- are protected from over-taining letdown pressure by automatic liquids with dis- controls and relief valves, solved gases therefore only minor leaks are considered possible.

The total gas content of the gas stripper and associated gas holding tanks is less than the holdup tanks in the Gaseous Waste Gas Disposal-System (Section 14.4.4) so even a total release via the Auxiliary Vent System could be accommodated (Section 9.13). Boron recovery i.eak Only degassed liquids are tanks normally stored in these tanks which are protected by dikes capable of retaining the entire content of the tank. The c;likes

are Class I structures
9.2.3-4 12-1-69 Component TABLE 9.2.3-1 (CONTINUED)

Malfunction Comments and Consequences Gas stripper and Fail to func- Letdown due to boration of the associated pumps tion Reactor Coolant System can be heater and con- diverted directly to the boron .trols recovery ~anks which are vented through the monitored Gaseous Waste Disposal System. Dilution letdown can be delayed. One boron recovery Fails to func- Degassed letdown is directed to the boron recovery tanks. Two evaporator or tion auxiliaries evaporators are provided so 50 percent of evaporator capacity will still be available while repairs are made. Sufficient capability to make boric acid solution for station requirements exists in the boric acid batch tanks, and the primary grade water tanks can supply adequate quantities of water. Primary grade Fails to func- Two 100 percent capacity pumps water pump tion are provided to permit main-tenance.

9.2.4-1 12-1-69

. 9.2.4 TESTS AND INSPECTIONS Periodic tests, calibrations, and checks are conducted on the various instrument channels to assure proper instrument response and operation of alarm functions.

Standby pumps are switched on a periodic basis, and continuously running equip-ment is inspected periodically to ensure availability. Routine inspections are performed on this system in accordance with maintenance procedures, and periodic tests are performed to ensure that standby equipment will perform in alternate failure conditions

FIG. 9.2- I H

OCT.15,1970 L 62-l/\L::l- 8v i71 I 2.'-CW-88-154-2~BR-IG,.0-152 A. FM-107F _G-"-SEOU5 'w~&Tt.__, 3'~GW* 21-1'.: 4 PRl~R'f' DRAIN ,-----'--""'--';;,t- G~C::.~OU5 """'S,TE (FIJl*\o7Bj CLIS~ TA.MK VE.NT c.HILLCfl C&i!OENS[R. __ c:;..,1,.sic.ous ,,11...*,>TE-, ~"":GW*l-15'2: ( FM*\078) 1'12*-cc-259-1:S 1 (FM*22D) 3°-BR-ltl-15'2 ti! 'j) l/DS*l5Y~ ii L.0. -.:..vos-1!-.Y BOff:lN RECOVEPY TAN i

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(FM*IO:.,Bj g!((-38*151 Lt JC I .

f-'c-!">*151 (FM*UD) I. VDS*ISY ~~~~tilJc. NO~TH Sl.)E. CF fl,.OAt> CL.15 2, C.L.154.

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ii 'JOS*I~

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4-Blt:36*152. TS-150

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FT /BR1"SO 0E.LECTR1C TO Pt,..IEUMA.TIC CONVE.'R.TE.fl. IIRl~O t13 AIR OPERATED DIAPHRAGM VALVE ' B~l:1,1 B:1~:c / HOT£7 ~FEU>JCE D\IIQ5:

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21A I 1 1-'t"'..BR-54-142. FLOW DIAE SORON RECOVERY SYS SHEET-'2,'3$4 FM*"Z9B,C ¢D FLOW DIA6 UQ WASTE DISPOSAL SYS SH'EET*l,"2~3 All*30A;8$C FLOW D~ FUEL PIT SYSTEM FM -~IA 1°-DA*l:?i.-151 FLOW 1'1,\6 SAMPLING SY,STEM ft,,11*:l'2.A FL.O'<N' Dliitl.~ VENT ¢ CRAIN SYSTEM F~-l~A FLOW 01.Mi REACTOR COOLANT SYS 11448 l'M*I03A.

FL.OW 1-.6 REACTOR COOLANT SY6 tlS48 Fl1-103A FLOW D1t.G CHEM,$ VOL CONT 5'f'6 su:r.1,2.{3 114.48 FM*J05A$B VDS-15'1' FLOW Dt"'6 CHEM$ VOL CONT SYS SIIEET-2 ltS4El, FH*IO"SB FL.OW' CllM SAF!:J'Y INJECTION SYSTEM fM*lOE,A FLOW DUrr6 GASE:JU 5 WASTE DISPOSAL SYST~M FM*I07i

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VEIJ>r,t~ VDS-ISY SVS (SM*-") F~M _ T~*BR.IO!I Lr~~- - -' T.5-15() \ [._l!!!..., SWITC,-: "

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3"-BR-15io-lS"2. c; BORON RECOVERY Plll~AR.Y OIUI.IN TANK PUMr.io OORON REC'JVERY FILTERS l SYSTEM-UNIT I-SH. I SURRY POWER STATION

FIG. 9.2-2 B H L OCT.15, 1970 86G-lN.:l-917iYl l 6MC:Ol15 WASTE-* J i° 3"-~ , ' ... ~-CW* 91 -154-

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BRIOIA. __§'j_ I* SR*E-76 BRI01B o,s:,::~:: ( j \ BRIOSB BR~2.~ jrlI 1*. BORON El/A.PORA.TOR LS p p~~;RHE.,b,.D COHOENSE.R IB~~B~~~;.-1 BORON EVAPORATOR BOTTOMS COOLER PRE!-lE:ATER

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01STILLl>.1"E :_ . r 1

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BOTTOMS COOLE.i..

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3. LOCATE ABOVE ElOT1"0M'C,1"At-lK

4. 30 l='EE.T OF Pl PE. 10 PITCH BACK TO TAN.K,NO INSYLATION.

TR 5 . ~ A P \ LLARY TUBING BliUi.O Co. JJ 0),,,,1 SORO~ COtJTROL PAW.EL. IZ!ELE.CTRIC TO PNEUMATIC CONVERTER

'I i

BOPON aeC::oYERY ~

SYSTEM

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t£*Bll-83-15'1 ..+/--~ BORON EVAPORATOR BOTTOMS PUMP5 BORON RECOVERY SYSTEM-UNIT I-SH. 2 SURRY POWER STATION

FIG.9.2'-3, OCT.15, 1970

H ' L

>62-~ .:1-8171711

                                                                                                                                                                                                                                                                              /Mlfi  l1

TEST TANKS PRIMARY WATER. TA.NU PCV RE.Sit.I WASTE.

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MIIIH - CATION UD FUEL PIT CESIUM lEMOVAL DtlOUTllJG lll'.MIICIALIZlla DEMl~tz.* ION El.awrl&Ell ION. DiGHANGERlil DEMINEI.AL~Elt$ H ...i*II 2*Ct+l*1 ....... 1-al-l t-c.H*J-2 l*CIH-1 1-Cltt* DS*l!i'f HC

            -K~-1)2.

2"*CH*t\Xl*l!4-5--'.::: (1104A*FM*IO!lil) . UQUID W6'&TE. l.52. BORON RECOVERY SYSTEM-UNIT I-SH. 3 SURRY POWER STATION

FIG. 9.2-4

OCT. 15, 1970 G6Z-lAJ..:l-917171 I CI.\EM.\C.A.L (;. 'IOLUME. CON'TROL SYSTEM

- HOSE COt,JW FOt. SAMPLINC. SYSTEM

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J; dl' 0- oJ BORON ION CLE.ANUP EXCHAt,IG.ER5 VC.S*IS"X:

CTYPi Hl&H LEVEL WASTE. DRAIN TA.t41( PUMPS (FM*~OA) IZ.E&IIJ T....a<. FILL

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VBS-ISY-VDS-ISY (TYP)

      +"'-F<H0-152 F"M-29C                                                                                                                                                            r~-PG*84*1S2 l!ORON A.£C.OVERY i-PG-S3*152 r-P6*"9-ISL
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UNIT I SAfEGUARDS. ARE.A i-CH-195-152 (FM-IOSB) ,-2."*CK-1~*151 P>--<<,..,......-(><:],-<', C~EH.1 CAL i VOLUME COWTRO\. ivsnM. TO WA&TE DISPOSAi. BORON RE.tOVE.RY E.IW'Ol<ATOll To BOR.O)r,J EV,._,._ 60Tl'OM5 TANK CIIZ.C PUMPS ~M-<S&) CIRC PUMP c1;~-~~P

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SAFEGUARDS ARLA 11*,lti!DE R£ACTOIZ C,Ol,.ITMl,IM£MT

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9.3.1-1 4-15-70 9.3 RESIDUAL HEAT REMOVAL SYSTEM

9. 3 .1 . DESIGN BASES The Residual Heat Removal System is shown in Figure 9.3-1. It is designed to remove residual and sensible heat from the core and reduce the temperature of the Reactor Coolant System during the second phase of unit cooldown. During the first phase of cooldown, the temperature of the Reactor Coolant System is reduced by_transferring heat trom the Reactor Coolant System to the Steam and Power Conversion System (Sectio11 10).

The Residual Heat Removal System is designed to be placed in operation when the reactor coolant temperature has been reduced to approximately 350°F and the reactor coolant pressure is between 400 and 450 psig. reactor shutdown. These conditio~s are assumed to occur approximately four hours after The system is designed to reduce the temperature of the Reactor Coolant System to 140°F within approximately 16 hours after the Residual Heat Removal System is placed in operation. The .residual heat generation at 20 hours after reactor shutdown, based on infinite core operation, *is approximately 66.0 x 106 Btu/hr. The system design precludes any significant reduction in the overall design reactor shutdown margin when the system is brought into operation for residual heat removal by equalizing t~e boron concentration and the temperature with the Reactor Coolant System

9
3 .1.~_2___ _

12-1-69 System components, whose design pressure and temperature are less than the Reactor Coolant System design limits, are provided with redundant isolation means and overpressure protective devices. A Residual Heat Removal System is provided for each unit. -Any leakage from the Residual Heat Removal System goes either to the containment or t~ the Component Cooling System which is a closed system. Any migration of radioactivity would be detected by the containment particulate and gas monitors (section 11.3) if the leak was to the containment. If the leak was to the Component Cooling System, the com-ponent coolingwat~r monitor would alarm in the event the radiation level reached a preset level above the normal background. All active system components which are relied upon to perform their function are redundant and the system design includes provisions to enable periodic hydrostatic testing to applicable code test pressures. Codes and Classifications All piping and components of the Residual Heat Removal System are designed to the applicable codes and standards listed in Table 9.3.1-1. SincE~ the system contains reactor coolant when it is in operation, austenitic stainless steel piping is used. The Residual Heat Removal Sy~tem is a seismi~ Class I system.

.7

9.3.1-3 12-1-69 TABLE 9.3.1-1 RESIDUAL HEAT REMOVAL SYSTEM CODE REQUIREMENTS Residual heat exchangers ASME III, Class C - Tube Side ASME VIII, Shell Side Residual heat removal piping and valves USAS B31.1 Residual heat removal pumps No Code

9 .. ~ l 4-15-70 - 9.3.2 SYSTEM DESIGN AND OPERATION 9.3.2.1 System Description The Residual Heat Removal System, shown in Figure 9.3-1 consists of two residual heat exchangers, two residual heat removal pumps and associated piping, valves and instrumentation. Two pumps and two residual heat exchangers perfonn the decay heat transfer functions for the reactor unit. After the Reactor Coolant System tempera-ture has been reduced to approximately 3S0°F and the reactor coolant pressure is between 400 and 450 psig, further system cooling is initiated .e by aligning the pumps to take suction from the reactor outlet line and discharge through the heat exchangers into the reactor inlet line. If only one pump and one heat exchanger are available, reduction of reactor coolant temperature is accomplished at a lower rate over a longer period of time. During unit cooldown, reactor coolant flows from the Reactor *coolant System to the residual heat removal pumps, through the tube side of the residual heat exchangers and back to the Reactor Coolant System. The inlet line to the Residual.Heat Removal System is located in the hot leg of reactor coolant loop number one between the main loop s~op valve and the reactor core. The return line connects to the cold legs of two loops through the Safety Injection System. The heat loads are transferred by the residual heat exchangers to the component cooling water in the Component Cooling System (Section 9.4).

9.3.2-2 12-1-69 During unit cooldown, the cooldown rate of the reactor coolant is controlled by regulating the flow through the tube side of the residual heat exchangers. A single bypass line and a remotely operated control valve around both residual heat exchangers are used to maintain a constant coolant flow through the Residual Heat Removal System while controlling coolant temperature. The entire Residual Heat Removal System is l o c a t e ~ t h e containment with the exception of the line penetrating the containment which connects to the refueling water storage tank. During refueling, the water level in the reae~r---cavity is lowered by opening a valve at the residual heat removal pump discharge and then pumping the water into the refueling water storage tank. The normal residual heat removal line is closed during the transferral. The Residual Heat Removal System is not an Engineered Safeguards System. 9.3.2.2 Components Residual Heat Removal System component design data are listed in Table 9.3.2-1.

9.3.2-3 6-30-71

TABLE 9.3.2-1 RESIDUAL HEAT REMOVAL SYSTEM COMPONENT DESIGN DATA Residual Heat Removal Pumps Quantity 2 Type Inline Centrifugal Capacity (each), gpm 4000 Head at rated capacity, ft H o 225 2

Motor horsepower, hp 300 Material Austenitic stainless steel and equivalent corrosion resistant

Design pressure, psig Design temperature, °F materials 600 400 Residual Heat Exchangers Quantity 2 Type Shell & U-tube 6

Design heat transfer rate(each), Btu/hr 33 X 10

9.3.2-4 12-1-69 TABLE 9.3.2-1 (Continued) _!~esidual Heat Exchangers (cont'd) Shell (component cooling water) Design temperature, °F 200 Design pressure, psig 150 Design flow rate, lb/hr 4.45 X 10.6 Design inlet temperature, °F 105 Design outlet temperature, °F 112 Material Carbon Steel Tube (Reactor Coolant) Design temperature, °F Design pressure, psig 400 600 6 Design flow rate, lb/hr 2 X 10 Design inlet temperature 140 Design outlet temperature 124 Material Austenitic Stainless Steel

9._3.2-5 12-1-*69 Residual Heat Exchangers

The residual heat exchangers are of the shell and U-tube type with the tubes welded to the tube sheet. Reactor coolant circulates through the tubes, while component cooling water circulates through the shell side.

The tubes and .other surfaces in contact with reactor coolant are austenitic stainless steel, and the shell is carbon steel. Residual Heat Removal Pumps The two 50% capacity residual heat removal pumps are in-line vertical centrifugal uni.ts with special seals to prevent reactor coolant leakage. All pump parts in contact with reactor coolant are austenitic stainless stee_l or adequate corrosion resistant material. Residual Heat Removal System Valves The valves used in the Residual Heat Removal System are constructed of austenitic stainless steel or other adequate corrosion resistant materia1s such as Haynes alloy 25 and 17-4 PH stainless steel. Manual stop valves are provided to isolate equipment for maintenance. Throttle valves are provided for remote and manual control of residual heat exchanger tube side flow. Check valves prevent reverse flow through the residual heat removal pumps

9.3.2-6 12-1-69 Isolation of the Residual Heat Removal System is achieved with two remotely operated stop valves in'series in the pipe from a reactor hot leg to the suction side of the residual heat removal pump, and by a check valve (located in the Safety Injection System) in series with a remotely operated stop valve in each line from the residual heat removal pump discharge on two reactor cold legs. Overpressure in the Residual Heat Removal System is relieved through a relief valve to the pressurizer relief tank in the Reactor Coolant System.

Valves that perform a modulating function are equipped with two sets of packing and an intermediate leakoff connection that discharges to the Vent and Drain System (section 9.7). Manually operated valves have backseats to faciliate repacking and to limit stem leakage when the valves are open. Leakoff connections are provided where required by valve size and fluid conditions. Residual Heat Removal Piping All Residual Heat Removal System piping_ is austenitic stainless steel. J>iping is welded except at the flanged connections of the flow control valves.

9.3.3-1 2-1-72

9. 3.3 SYSTEM DESIGN EVALUATION 9.3.3.1 Availability and Reliability For Reactor Coolant System cooldown, the Residual Heat Removal System is provided with two pumps and two residual heat exchangers, If one of the two pumps and/or one of the two heat exchangers is not operative, safe operation of the unit is not affected; however, the time for cooldown is extended.

9.3.3.2 Incident Control The suction side of the Residual Heat Removal System is connected to the

-    reactor coolant hot leg of one loop and the discharge side to the cold legs of the other two loops through the Safety Injection System. On the suction side the connection is through two electric motor-operated gate valves in series. The first valve is interlocked with Reactor Coolant System pressure so that if the Reactor Coolant System pressure exceeds a set pressure, the valve does not open. The second valve, prior to unit startup after the first refueling, will be similarly interlocked. In the interim, electric power to this valve will be locked out with the valve in the closed position whenever the reactor coolant pressure exceeds 465 psig. On the discharge side of the Residual Heat Removal System, each connection is made through an electric motor-operated valve in series with a check valve. All of these valves are remote manually*closed whenever the Reactor Coolant System pressure and tempera-ture exceed approximately 450 psig and 350°F, respectively.

iA.,,

9.3.3-2 4-15-70 The fluid operating pressure is higher at all times on the tube side of the residual heat exchanger than on the shell side, ranging over an approximate range of 450 to 100 psig, so that in case of leakage, reactor coolant leaks into the component cooling water in the shell side. Abnormally high temperature of the component cooling water would be indicated the Main Control Room, at which time the control valve in the vent line from the component cooling surge tank to the process vent would be closed by manual operation of a push button in the Main Control Room, if it had not previously closed automatically due to high radiation signals from transmitters installed in the component cooling water piping. Inleakage to the component cooling water, if not stopped, results in high level in the com-ponent cooling surge tank, and eventually fills the tank. Excess water from the tank is disposed of by a relief valve discharging to the auxiliary building sump. After the leakage condition is corrected, radioactivity in the component cooling water is reduced by bleed and feed. The residual heat removal pumps are driven by drip-proof type motors

                                                                                 -    J' with Class B epoxy type insulation to be capable of operation in high humidity conditions. They are equipped with splash barriers to protect the motors 1.n the event of a pipeline break in the area, which could possibly spray and wet the ltl)tors.

The inlet line from the Reactor Coolant System to the Residual Heat Removal System is between the reactor core and the outlet loop isolation valve. Thus, if the outlet or inlet loop isolation valve is closed, the inlet from the Reactor Coolant System to the Residual Heat Removal System is not (

I blocked.

9.3.3-3 12-1-69

9.3.3.3 Malfunction Analysis A fai-lure analysis of residual heat removal pumps, heat exchangers and valves is presented in Table 9.3.3-1
9.3.3-4 12-1-69 TABLE 9.3.3-1 RESIDUAL HEAT REMOVAL LOOP MALFUNCTION .ANALYSIS Component Malfunction Comments and Consequences

1. Residual heat Rupture of The casing and shell are designed removal pump casing for 600 psig and 400°F. The pump is protected from overpressurization by a relief valve in the piping discharging to the pressurizer relief uank. The pump can be inspected and is located in the containment structure.with protection against missiles.

considered credible. Rupture is not

2. Residual heat Pump fails One operating pump furnishes half removal pump to start of the flow required to meet design cool-down rate. This only increases the time necessary for unit cooldown.

3. Residual heat Manual valve This is prevented by administrative controls removal pump on pump during prestartup and operational check.

suction is closed

9.3~3-5 12-1-69 TABLE 9.3.3-1 (Continued) Component Malfunction Comments and Consequences

4. Residual heat Stop valve in Prestartup and operation checks removal pump discharge confirm position of valves.*

line closed or check valve sticks closed

5. Remote Valve fails Valve position indication light operated valve to open indicates that the valve has not opened, inside con- Valve is opened manually or unit tainment in is slowly cooled by feed and. bleed

                                                         ---*----~--------

pump suction procedures. line

6. Remote Valve fails Two valves in parallel. If one valve operated valve to open fail$ to open, flow passes through

     *inside con-                    other valve.

tainment in pump dis-charge line

9.3.3-6 12-1-69 Component TABLE 9.3.3-1 (Continued) Malfunction Comments and Cons~quences

7. Residual heat Tube or Rupture is considered very unlikely exchanger shell because of low operating rupture pressure as compared to design pressure. In any event the faulty heat exchanger can be isolated and the remaining heat exchanger used for cooldown. With only one heat exchanger the time for cooldown is extended.

8. Valve in Valve sticks Part of flow does not pass through bypass line open residual heat exchangers. This around increases the time for unit residual heat cooldown.

exchangers

9.3 4-1 12-i-69

9.3.4 TESTS AND INSPECTIONS The residual heat removal pump flow instrument channels are calibrated during each refueling operation.

The active components of the Residual Heat Removal System are in inter-mittent use during normal plant operation and no additional periodic tests are required. Periodic visual inspections and preventative maintenance are conducted following normal industrial practice. Samples are analyzed to determine the amount of radioactivity in the reactor coolant. If the radioactivity level is high, a reactor coolant sample is analyzed and an iodine extraction made and counted as an indication of

defects in fuel cladding. The frequency of sampling for gross activity and for radiochemical analysis of the reactor coolant will be specified in the Technical Specifications.

With the exception of the line penetrating t~,~--_,containment_a11.c:l going to

              -------                      .      ..   -~                                     ... *.        .     *-- ........,__,,

the--refueling water storage tank, the entire Residual Heat Removal Sy.stem (

/ /

is located inside the containment. ......-....... -

                                                                                                                                    /!
                                                                                                                                      \
     ~-------**..*-*******-*--*****-----*****-****'---*-** --*- .... ----*-**** -
                                     '                                            -*~>~, ... .,...-~*<*~r,~

FIG:9.3-1 OCT.15,1970 F

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                                                                                                                                                                                                                                                                                                                                                     'f . PLUGGED CONNECTION I

RESIDUAL HEAT REMOVAL SYSTEM SURRY POWER STATION

9.4-1 12-1-69 9.4 COMPONENT COOLING SYSTEM The Component Cooling System consists of the Compone~t Cooliri13 Water, the S!Jilled Component Cooling Wat;r, Chilled Water, Neutron Shield Tank Cooling

 . Water, and ~harging Pump Coolini Water Subs~stems. These subsystems are used individually or in combination wi,th each other to provide cooling water for removal of heat from components in the station. The Component Cooling System including subsystems is shown on Figures 9.4-1 to 9.4-9.

Component Cooling Water Subsystem The Component Cooling Water Subsystem, serving two reactor units, is designed

~

to: Provide cooling water for the removal of residual and sensible heat from the Reactor Coolant Systems durin~ unit shutdown and cooldown.

2. Cool the spent fuP.l pit water.

Cool the reactor coolant pump motor coolers.

4. Cool the letdown flow in the Chemical and,Volurne Control Systems during power operation.

{I) ~ the reactor coolant pump seal water return flow.

9.4-2 2-15-71

6. Normally supply cooling water for the containment recirculation air coolers and neutron shield tank coolers.

71* Provide makeup water for the Neutron Shield Tank Cooling Subsystem

(: ~'

  @        Provide makeup water for the Charging Pump Cooling Water Subsystem *
 .~        Provide cooling to dissipate waste heat from other reactor and station components.

Chilled Component Cooling Water Subsystem J ~ Chilled Component Cooling Water Subsystem is provided for both reactor units and is designed to provide low temperature component cooling water to the con-tainment recirculation air coolers, the neutron shield tank coolers, the primary drain tank vent chiller condenser and the waste gas recombiner after-Gooler when the component cooling water tempergture exceeds 75° F (See Section 9.4.3.2). Chilled Water Subsyster.. A separate Chilled Water Subsystem is provided for each reactor unit~ Each Chilled Water Subsystem is designed to provide chilled water for use in cooling the chilled component cooling water. Each Chilled Water Subsystem is also

      ~~pable of cooling the water in the refueling water storage tank after a
    .refueling operation.

9.4.1-1 2-13-70 9.4.1 DESIGN BASES 9.4.1.1 Component Cooling Water Subsystem The Component Cooling Water Subsystem is an intermediate cooling system and transfers heat from heat exchangers containing reactor coolant or other radio-active liquids to the Service Water System (Section 9.9). The maximum heat load occurs during the initial stages of residual heat removal during a reactor unit cooldown. The Component Cooling Water Subsystem is designed to reduce the temperature of the reactor coolant to 140° F within 20 hours after a reactor shutdown based on river water temperature of 95° F

During normal full-power operation, one component cooling pump and one component cooling heat exchanger acco!!'lm.adates the heat removal loads for each reactor unit.

Operation of two pumps and two heat exchangers is the standard procedure during the removal of residual and sensible heat during unit cooldown, although one pump and one exchanger may be safely used under these conditions. Each of the four component cooling heat exchangers is designed to remove the entire heat load from one unit plus half of the heat load common to both units during normal operation. Each heat exchanger is also capable of removing half of the heat removal load occurring 4 hours after a shutdown of one unit under conditions representing the maximum allowable cooldown rate. The presence of excess radioactivity in the Component Cooling Water Subsystem is detected by two gamma scintillation radiation monitors. High radiation signals

9.4.1-2 12-1-69 from either of these detectors cause the surge tank vent isolation valve to shut and initiate an alarm in the Main Control Room. One detector monitors the supply to Unit 1, and is mounted on the 18 in. combined discharge line from com-ponent cooling heat exchangers 1-CC-E-lA and-lB. The second detector monitors the supply to Unit 2, and is mounted on the 18 in. combined discharge line from component cooling heat exchangers 1-CC-E-lC and -lD. Both detectors are located in the Unit 1 turbine ronm to prevent possible interference from back-ground radiation levels in the auxiliary building. Operation of these detectors fs described in Section 11. 3. 3 Component Cooling Water Subsystem component design data are given in Tabie 9.4-1. The Component Cooling Water Subsystem is designed as Class I (Section 15.2.1). 9.4.1.2 Chilled Component Cooling Water Subsystem The Chilled Component Cooling Water Subsystem is designed to provide 70° F chilled component cooling water to the containment recirculation air coolers, the neutron shield tank coolers, the primary drain tank vent chiller condenser and the waste gas recombiner aftercooler as nece~sary (See Section 9.4.3.2). Three chilled component .coolers and taree chilled component cooling pumps are provided for the Chilled Component Cooling Water Subsystem that serves both reactor units. I

9.4.1-3 12-1-69

A pump and cooler serves each reactor unit and one pump and cooler is provided for use as a spare.

The piping is arranged so that the spare cooler and pump can be used together for either reactor unit or individually to replace a.component normally used, for either reactor unit. Chilled Component Cooling Water Subsystem component design data are given in Table 9.4-3. 9.4.1.3 Chilled Water Subsystem

*Each Chilled Water Subsystem is designed to provide 350 tons of refrigeration at 60° F water temperature. This cooling capacity is sufficient to cool and maintain the Chill_ed Component Cooling Water Subsystem at 70° F when the Chilled Component Cooling Water Subsystem is operating at maximum capacity.

Chilled water is also used directly to cool the 350,000 gal of water in the refueling water sto,rage tank from 100° F to 40° F in 4 days after a refueling operation if necessary. Cooling the water in the refueling water storage tank does not effect the ability of the Chilled Water Subsystem to cool the Chilled Component Cooling Water Subsystem because the refueling water storage tank cooling operation is performed prior to reactor startup. A chilled water unit, two chilled water circulating pumps, and a chilled water ~ condenser are provided for the Chilled Water Subsystem of each reactor unit.

9.4.1-4 12-1-69 I The chilled water unit has two 100 percent capacity ejectors. Two 100 percent capacity chilled water circulating pumps are provided to assure sufficient spare capacity so that failure of _one component will not render the subsystem inoperable. Chilled Water Subsystem component design data are given in Table 9.4-2. 9.4.1.4 Neutron Shield Tank Cooling Water Subsystem The Neutron Shield Tank Cooling Water Subsystem is designed to remove heat from the neutron shield tank. Two neutron shield tank coolers, a neutron shield surge tank, a corrosion control tank, and all necessary piping and valves the subsystem serving a reactor unit. Each neutron shield tank cooler is a 100 percent capacity cooler. The second cooler is a spare which can be placed in operation remotely by means of motor operated valves. Neutron Shield Tank Cooling Subsystem component design data are given in* Table 9.4-4. 9.4.1.5 Charging Pump Cooling Water Subsystem The Charging Pump Cooling Water Subsystem is designed to:

1. Provide cooling wate.r to the charging pump mechanical seal heat coolers

9.4.1-.5 12-1-69

2. Provide cooling water from the Service Water System to the charging pump intermediate seal coolers and to the charging pump lubricating oil coolers Charging Pump Cooling Water Subsystem component design data are given in Table 9. 4-5.

The Charging Pump Cooling Water Subsystem is designed to the Class I earthqu~ke criteria

9.4.1-6 12-1-69 TABLE 9.4-1 COMPONENT COOLING WATER SUBSYSTEM COMPONENT DESIGN DATA Pumps Number 4 (2 required for normal operation of 2 reactor units)

Type Horizontal centrifugal single stage Motor horsepower, hp . 600 Seal Single mechanical Capacity, gpm 9,000 Head at rated capacity, ft 200 Design pressure, psig 200 Design temperature, °F 220 Materials Pump casing Shaft Cast iron Alloy steel, ASTM Al07, Grade 1045 Impeller Cast iron

9.4.1-7 12-1-69

Heat Exchangers TABLE 9.4-1 (Cont'd)

Number 4 (2 required for -normal operation of 2 reactor units) Duty, Each, Btu/hr 50 *.J X 10 6 Shell Tube Design pressure, psig 150 150 Design temperature, OF 150 150 Operating pressure, psig 95 5.6 Operating temperature, in/out,°F 119. 7-105.0 95.0-106.2 Materials Carbon .steel 90-10 Cu-Ni Fluids Component Service water cooling water Design code ASME III ASME III Class C Class C

. I

9.4.1-8 12-1-69 TABLE 9.4-1 (Cont'd)

Surge Tank Number 1 (common to both units)

Type Cylindrical, horizontal Capacity, gal 2,810 Design pressure, psig 40 Design temperature, °F 150 Material Carbon steel Design code ASME III Class C Chemical Addition Tank Nwnber 1 (conunon to both units) Type Cylindrical, vertical Capacity, gal 120 Design pressure, psig 150 Design temperature, °F 150 Material Carbon steel Design code ASME VIII

9.4.2-1 12-1-69

9.4.2 PIPING AND VALVES (CHECK AND MANUALLY OPERATED GATE, BUTTERFLY AND GLOBE)

Carbon steel pipe is used throughout the system, and joints are welded*except where flanges are used at connections to equipment and to butterfly and check valves in sizes 10 in. and larger. All valves are of steel material-except the butterfly type, which are cast iron. Selected piping, valves, and supports. are designed as Class 2. Expansion joints are provided at the suction and dis-charge of the component cooling water pumps. The piping system conforms to the requirements of the USA Code for Pressure Piping B-31.1. Relief Valves

Small thermal relief valves are constructed with stainless steel body and trim and carbon steel bonnet and cap.

with stainless steel trim

Larger relief valves have carbon steel body
)

9.4.2-2 12-1-69 TABLE 9.4-2 CHILLED WATER SUBSYSTEM COMPONENT DESIGN DATA Chilled Water Condenser Number 2 (one for each unit) Type Shell and tube Duty, Each, Btu/hr 16,217,400 Steam condensed, lb/hr 15,000 Circulating water flow, gpm 2,150 Waterbox design pressure, psig 50 Circulating water inlet/outlet temperature, oy 95/ 110 Materials Shell Water box Tubes Carbon steel Carbon steel lined with Carbonmaster 14 90-10 Cu-Ni

9.4.2-3 12-1-69

TABLE 9.4-2 (Cont'd)

Chilled Water Units Number 2 (one for each unit) Flash tank design temperature, °F 300 Flash tank material Carbon steel Flash tank design pressure Full vacuum Ejectors 4 (two per chilled water unit) Steam pressure required, psig 110 Steam flow per ejector, lb/hr 11,000 Capacity, tons of refrigeration 60° F water 350 per unit

38° F water Chilled Water Circulating Pumps 125 per unit Number 4 (two for each unit)

Type Horizontal centrifugal, single stage Motor horsepower, hp 90.6 Seal Single mechanical Capacity, gpm 1,200 Head at rated capacity, ft 299.1 De~ign pr~ssure, psig 300 Design temperature, °F 250 Materials Pump casing Cast iron

Shaft Impeller Alloy steel, AISA C-1045 Bronze

9.4.2-4 12-1-69 TABLE 9.4-3 CHILLED COMPONENT COOLING WATER SUBSYSTEM COMPONENT DESIGN DATA Chilled Component Cooler Number 3 ( one for each unit, one common to both units) Duty, Each, Btu/hr 3,600,000 Shell Tube Design pressure, psig 150 150 Design temperature, °F 150 -150 Operating pressure, psig 60 60 Operating temperature, in/out, °F 80/70 60/66 Materials Fluids Design code Carbon steel Component cooling water ASME III Admiralty Chilled water ASME III Class C Class C

9.4.2-5 12-1-69

TABLE 9.4-3 (Cont'd)

Chilled Component Cooling Pumps Number 3 ( one for each unit, one common to both units) Type Horizontal centrifugal, single stage Motor horsepow.er, hp 34.2 Seal Mechanical Head at rated capacity, ft 187.5 Design pressure, psig 250 Design temperature, °F 250 Materials Pump casing Cast iron Shaft Alloy steel, AISA C-1045 Impeller Cast iron

9.4.2-6 12-1-69 TABLE 9.4-4 NEUTRON SHIELD TANK COOLING SUBSYSTEM COMPONENT DESIGN DATA Neutron Shield Tank Cooler Number 4 (2 for each unit, one required) Duty, Each, Btu/hr 80,000 Shell Tube Design pressure, psig 150 50 Design temperature, °F 100 150 Operating pressure, psig 50 15 Operating temperature, in/out, °F 80/85 125/90 Materials

Fluids Design code Type 316 Component ss cooling water ASME Section VIII Type 316 ss Shield tank water ASME Section VIII

9,.4*.2,-7 12:..1-69*

TABLE 9.4-4 (Cont'd)

Nuetron Shield Tank Surge Tank Number 2 (one for each unit) Type Cylindrical, vertical Capacity, gal 1,444 Design pressure, psig 25 o* Design temperature, F 150 Material Carbon steel Design code ASME VIII Corrosion Control Tank

Number Type Capacity, gal 2 (one £'or each unit)

Cylindrical'., vertical 158 Design. pressure, psig Il50 0 Design- temperature, F 150 Material Stainless steel, Type 304 Design code ASME VIII

9.4.2-8 12-1-69 TABLE 9.4-5 CHARGING PUMP COOLING WATER SUBSYSTEM COMPONENT DESIGN DATA Charging Pump Cooling Water Pump Number 2 per unit Type Centrifugal, inline, single stage Motor horsepower, hp 5 Seal Single mechanical Capacity, gpm 90 Head at rated capacity, ft 105 Design pressure, psig 150 Design temperature, °F 250 Materials Pump casing Shaft Cast iron Alloy steel Impeller Bronze

9.4.2-9 12-1-69,

TABLE 9.4-5 (Cont'd)

Charging Pump Service Water Pwnp Nwnber 2 per unit Type Centrifugal, inline, single stage Motor horsepower 3 Seal Packing Capacity, gpm 90 Head at rated capacity, ft 55 Design pressure, psig 150 Design temperature, °F 250 Materials Pmnp casipg Cast iron Shaft Alloy steel Impeller Cast iron

9.4.2-10 12-1-69 TABLE 9.4-5 (Cont'd) Charging Pump Intermediate Seal Cooler

   'Number                                  2 per unit Duty, Each, Btu/hr                      45,600 Shell             Tube Design pressure, psig                     50               75 Design temperature, °F                  150               150 Operating pressure, psig                  25               40 Operating temperature, in/out, °F       95/97             106/ 105 Materials                               Bronze            Admiralty Fluids                                  Service water     Component cooling water Design code                             ASME Section VIII ASME Section VIII Charging Pump Seal Cooling Surge Tank Number                                  1 (common to both units)

Type Cylindrical horizontal Capacity, gal 20 Design pressure, psig Atmospheric Design temperature, °F 150 Mater:f,.al Carbon steel Design code ASME VIII

9.4-3 12-1-69

Neutron Shield Tank Cooling Water Subsystem The Neutron Shield Tank Cooling Water Subsystem is designed to circulate and cool the water in the neutron shield tank which is heated by neutron and gamma radiation.

Charging Pump Cooling Water Subsystem A separate Charging Pump Cooling Water Subsystem is provided for each reactor unit, and is designed to provide component cooling water for each charging pump mechanical seal cooler and service water to each charging pump lubricating oil cooler

9.4.3-1 12-1-69
9.4.3 9.4.

3.1 DESCRIPTION

Component Cooling Water Subsystem During operation, component cooling water is pumped through the shell side of the component cooling water heat ,exchangers, where it is cooled by service (river) water, and then through parallel circuits to cool the following components:

1. Reactor coolant pump thermal barriers~ bearing oil coolers, and motor stators

2.

3. Excess letdown heat exchangers "(intermittent heat load) Non-regenerative heat exchangers

4. Reactor unit and steam generator blowdown sample coolers (intermittent heat load)

5. Seal water heat exchangers

6. Residual heat removal pumps seal coolers and stuffing box jackets (during unit cooldown)

,. 7. Residual heat removal exchangers (during unit cooldown)

9.4.3-2 12-1-69 8. 9. Boron Recovery System equipment (intermittent heat load) Containment penetration cooling coils

10. Fuel pit coolers

11. Reactor shroud cooling coils

12. Primary shield penetration cooling coils

13. Primary shield water wall coolers 14.

15. Primary drain coolers Liquid Waste Disposal System equipment (intermittent heat load)

16. Gaseous Waste Disposal System equipment

17. Neutron shield tank coolers

18. Reactor containment air recirculation coolers The Component Cooling Water Subsystem is designed as a closed system, with a surge tank at the pump suctions. The tank is th~ high point of the system and provides the required net positive suction head for proper operation of the pumps. Th~ heat exchangers are located in the turbine room for Unit 1.

9. 4. 3.:.3 12-1-69

Pumps, tanks and some of the equipment cooled by the system are installed in the auxiliary building; the remainder of the equipment served is located in the reactor containments. Two 18 in. main supply and two 18 in. main return lines are used for each reactor unit. These mains, in full size, are connected directly to the residual heat removal exchangers which are located in the reactor containments and are at the extremities of the two piping loops.

Reduced size branches connected to the mains form cross circuits which serve the remainder of the apparatus being cooled. Equipment which is common to both reactor units is located in the auxiliary building, and associated cross circuits are double connected to the mains fi:>r both reactor units. High point vents and low point drains are provided as required by piping configuration.

Each cooling water outlet line from a piece of e,quipment contains a valve for controlling flow; the valve is either a manually operated globe type or an automatic air operated type positioned by pressure or temperature control signals originating in cooled systems.

The system is provided with trip valves for isolating the containment structures in accordance with the requirement of Containment Isolation System (Section 5.2). The system is monitored from the Main Control Room by indicators which dis-play the following data:

1. Pump discharge pressure

2. Radioactivity, .temperature, and flow in the supply mains immediately downstream from the heat exchangers.

9.4.3-4 12-1-69

3. Temperature and flow in the return mains at exit from the reactor containments.

4. Temperature in the return mains at the pump suctions.

5. Level in the surge tank.

6. Data which are of a common nature are displayed on both main control boards.

Pressure switches for automatic starting of standby pumps are installed in the pump discharge mains. Local indicators for pressure, temperature, level, and flow are provided on a general basis. Certain selected temperatures are sensed by thermocouples whose output signals.are fed into the Computer Monitoring System (Section 7.9) thus providing full time scanning and alarming. These temperatures can be read out during periods of abnormal values. Other important temperatures, as well as important pressures, levels, and flows are alarmed in the Main Control Room when abnormal values are reached. Thermal relief valves are installed around all equipment which might be over-pressured by a combination of closed component cooling water inlet and outlet valves and heat input from the isolated equipment. A relief valve, and a vacuum breaker valve, is provided for.the surge tank. Also, a relief valve is installed directly downstream from the level controlled valve supplying makeup to the tank. The surge tank level is automatically controlled at 12 in. above the centerline. Capacity above this level is 910 gal, which is sufficient to accommodate

9.4.3-5 12-1-69

minor system surges and thermal swell due to cooldown operation without over-flowing through the relief valve. Makeup from the turbine room Main Condensate System is admitted automatically through an air operated valve controlled from tank le,,e1. The makeup line is double connected to both Main Condensate Systems; this provides redundancy since it is unlikely that both turbine generators would be out of service during cooldown of a reactor unit. High level in the tank is lowered by manual operation of system low point drains. The tank is equipped with a full length gage glass and the makeup valve is provided with a handjack for local manual positioning of the valve in emergencies.

A 120 gal capacity chemical addition tank is connected to the pump suction and discharge piping. In operation, the tank is charged with chemicals after being

isolated and having its level lowered by manual valves. After closure of the charging and manual level valves and opening of the isolation valves, discharge pressure forces water into the tank and injects the mixture into the system at the pump suctions.

The desired water chemistry is obtained by the addition of potassium chromate for corrosion inhibition and potassium hydroxide for pH control. The design objective of the chemical treatment is to maintain 2,000 ppm of chromate, at 8-11 pH. Sampling is performed at the central station in the auxiliary building. Several local sample points are also provided. 9.4.3.2 Chilled Component Cooling Water Subsystem ~ The Chilled Component Cooling Water Subsystem provides cooling to the neutron shield tank, the containment recirculation air coolers, the primary drain tank

9, 4. *3-6 12-1-69 vent chiller-condenser, and the waste gas recombiner aftercooler whenever the component cooling water temperature rises above 75° F. One chilled component cooling water pump circulates water through a chilled component cooler where the chilled component cooling water is cooled to 70° F. When the Chilled Component Cdoling Water Subsystem is not required the subsystem is isolated from the Component Cooling Water Subsystem by a valve arrangement, The isolated section remains filled with water so that it can be operated without delay. When the component coolirig water temperature rises above 75° F, the equipment which requires 75° F cooling water is changed from the component cooling water header to the Chilled Component Cooling Water Subsystem by manually operating transfer valves for each piece of *equipment. Transfer valves for equipment located within the containment are operated remotely from the Main Control Room. Transfer valves for equipment located outside the containment are operated locally. 9.4.3.3 Chilled Water Subsystem A Chilled Water Subsystem is pr:ovi.ded for each reactor unit to *supply chilled water to the chilled component coolers or the refueling water storage tank coolers, Chilled water is produced in the chilled water unit which is a flash tank with two steam ejectors. Auxiliary steam at 150 psig is used to operate the ejectors. Each ejector is full capacity with one serving as a spare. The steam ejector produces a vacuum in the tank which causes a portion of the water inside to flash and lower the temperature of the remaining water. The ejectors discharge to the chilled water condensers where the steam is condensed. The chilled water in the flash tank is circulated through the chilled component coolers or the refueling water storage tank coolers by one of the two full size

9.4.3-7 2 71

chilled water circulating pumps. The chilled water returns to the flash tank at an elevated temperature.

Makeup for the Chilled Water Subsystem is from the Main Condensate System. Cooling water for the chilled water condenser is service water from the service water pump discharge. Controls are provided to maintain level in the flash tank and to prevent freezing of the chilled water. 9.4.3.4 Neutron Shield Tank Cooling Subsystem A Neutron Shield Tank Cooling Subsystem is provided for each reactor unit to cool the water in the neutron shield tank which is heated by neutron and gamma radiation from the reactor. The heated water in the neutron shield tank rises

by natural convection to the top of the tank and into the pipe connected to the neutron shield tank cooler. The cool water from the Component Cooling Water Sub-system or the Chilled Component Cooling Water Subsystem circulates through the neutron shield tank cooler, cooling the heated neutron shield tank water. Only one neutron shield tank cooler is required to perform the required cooling, the other cooler is a spare and is isolated from the subsystem by motor operated valves. A surge tank accommodates thermal expansion of the neutron.shield water.

A level sensor on the surge tank sends a signal to the Main Control Room to in~icate low subsystem level. A solenoid operated valve is actuated from the Main Control Room to replenish the subsystem from the Component Cooling Water Subsystem. The corrosion control tank is used for the addition of an inhihitor when the reactor is not operating; this is a manual operation

9.4.3-8 12-1-69 9.4.3.5 Charging Pump Cooling Water Subsystem A ~barging Pump Cooling Water Subsystem for each reactor unit provides com-ponent cooling water for the charging pump mechanical seal heat exchangers which cool the water circulating in the charging pump mechanical seal cooling loops, an-a service water to cool the charging pump intermedi_ate seal coolers and the charging pump lubricating oil coolers.

Either of two 100 percent capacity charging pump cooling water pumps circulates t~e component cooling water in the subsystem. A surge tank acconunodates thermal e~~ansion of the component cooling water. A level sensor in the surge tank a~tomatically actuates a makeup ~alve to replenish the subsystem from the Component Cooling Water Subsystem *. To ensure that component cooling water is c~~.tinually available to the mechanical seal coolers, one pump is in operation and the other pump on standby. The standby pump is automatically actuated on low pump discharge pressure to supply cooling water in the*event of failure of the operating pump. Two 100 percent capacity charging pump interme.diate seal co_olers are provided to cool the component cooling water which is circulated to th~. mechanical seal coolers.

 'J ~-*

Either of two 100 percent capacity charging pump service water pumps deliver water from the Service Water System to the charging pump intermediate seal coq.~ers and the charging pump lubricating oil coolers, thereby maintaining the charging pump lubricating oil and the component cooling water used to cool the charging pump mechanical seals at the proper temperature. To ensure that service water is continually available, one pump is in operation and the other

9. ,~. 3-9 10-15-70 on standby. The standby pump is automatically actuated on low pump discharge pressure to supply service water in the event of failure of the operating pump.

The installation of two full capacity charging pump cooling water pumps, two full capacity charging pump service water pumps, and two full capacity charg-ing pump intermediate seal coolers provides 100 percent redundancy for this cooling water subsystem. All components of the Charging pump Cooling water Subsystem, including pumps, heat exchangers, and tanks are designed to the Class~ earthquake criteria. Both the charging pump cooling water pump and the charging pump service water

pump are connected to the emergency electrical bus to insure that they will operate in the event of a loss of station power .

9.4.4-1 12-1-69

9.4.4 9.4.4.1 DESIGN EVALUATION Component Cooling Water Availability and Reliability The Component Cooling Water Subsystem uses machinery and equipment of conven-tional and proven design. All components are specified to provid~ maximum economy, safety and reliability.

The installation of four pumps and four heat exchangers for two reactor units provides 100 percent backup during normal operation of the two units. During I

                                                                                   'I cooldown of one reactor unit, there is 100 percent backup for it if the other unit is out of service and 50 percent backup if the other unit is in normal operation. If only one pump is available for cooldown of a reactor unit, the*

cooldown time is extended without equipment damage or hazard to the public or operating personnel *

  .Most of the piping, valves and instrumentation in the reactor containment are located outside the reactor primary shield wall and above the water level in the bottom of the containment under post-accident conditions. The exceptions are the lines for the neutron shield tank coolers and the primary shield penetration and water wall-cooling coils; these lines can be secured by valves located outside of the crane support concrete wall. The equipment in the containment is pro-:

tected against credible missiles and flooding during post-accident operations. Also, shielding is provided to allow limited maintenance and inspection during power operation.

9.4.4-2 12-1-69 Equipment not located in the containment may be inspected and maintained during power operation. The system is of Class I design and is designed to the codes stated in Section 9.4.1. The main piping loops and the loop for the fuel pit coolers are Class I category and are designed accordingly. Also, these loops are analyzed artd desig~ed to meet associated thermal stress requirements. The following components are located inside the containment: the exc~ss letdown heat exchanger; reactor coolant pump thermal barrier; oil coolers and motor stators; primary shield penetration and water wall coolers; neutron shield tank coolers, reactor shroud cooling coils; primary drain coolers; residual heat removal exchangers; containment air recirculation coolers; residual heat removal pump seal coolers; and pipe penetration cooling coils. Isolation of flow from the Component Cooling Water Subsystem to the containment is provided as described in Section 5.2. The component cooling surge tank, which normally operates at atmospheri.c pressure, is equipped with a vent line connected into the process vent system. The tank vent line contains an automatic shutoff valve; this valve, normally open, closes automatically upon receiving a high radiation signal from either of the two radiation monitors located on the discharge piping from the component cooling water heat exchangers, and can be manually closed from the Main Control Room. The high radiation condition which caused the valve closure is indicated by an alarm.

9.li.4-3 12-1-69

An air operated trip valve is installed in the outlet cooling water header from the reactor coolant pump thermal barriers, in the outlet aooling water line from the excess letdown heat exchanger, and in the outlet cooling water line from the primary drain cooler. A check valve is installed in the inlet cooling water header to the thermal barriers, bea,,ring oil coolers, stator coolers, and in the inlet cooling water line to the *excess letdown heat exchanger. In the event that a leak occurs in the thermal barrier cooling coil, an alarm annunciates in the Main Control Room and the high pressure reactor coolant is safely contained by*closing the appropriate stop valve. A high cooling water outlet flow signal from either the thermal barrier cooling header, the excess letdown heat exchanger, or the primary drain cooler automatically closes the isolation valves. The air operated stop valves in the outlet cooling water header from the thermal barriers, in the main cooling water lines leaving the reactor containment, and in the r~actor containment air recirculation cooler outlet lines leaving the reactor containment close on high containment pressure.

Leaka~e Provisions The component cooling water heat exchangers are located in the turbine building. Provisions are made to preclude the possible spread of radioactive contamination ev.en though this system is not normally expected to contain ra~ioactive water. Th~se precautions include isolation of each heat exchanger by manual shutoff of the inlet and outlet component cooling water valves, treating any leakage and water samples from these heat exchangers as radioactive, and installation of the heat exchangers within a curbed area to preclude radioactive contamination

- of the turbine building floor. Any leakage*is then returned to the Liquid Waste Disposal System (Section 11.2) via the sump pump provided within the curbed area.

9.4.4-4 10-15-70 Welded*construction is used almost exclusively throughout the system to minimize possibility of leakage from pipe, valves, and fittings. Small leakage inside the containment is not considered to be objectionable. Con-tamination could result from the following:- side-to-side leakage in a heat exchanger in the Chemical and Volume Control, Residual Heat Removal, or Sampling Systems, or a leak in the thermal barrier of a reactor coolant pump. Leakage from the system is primarily detected by falling surge tank level. Temperature, level, and flow indicators in the Main Control Room may be used to detect leakage at certain points . Elsewhere, leaks can be located by inspection or isolation. Incident Control The piping mains have the following valves at the containment walls: shutoff on outside and check on inside in supply lines; trip and shutoff on outside in return lines. The trip valves close upon receiving the containment isolation signal. Piping for the reactor coolant pumps, reactor shroud cooling coils, and containment air recirculation coolers is valved in an identical manner. See Section 5.2.2. The temperature of cooling water supplied to the reactor containment air recircu-o . lation coolers should not exceed 70 F; this means that during periods of warmer river water, these coolers use chilled component* cooling water. The cooler inlet and outlet lines are sized large enough so that abnormally high flows of regular (unchilled) component cooling water can be accommodated with *the globe valves wide op~n. With this added flow, the coolers can control containment pressure

9.4.4-5 12-1-6-9

- rise caused by a minor incident or pipe break, thus, avoiding use of the Safeguards System (Section 6.0) in this situation. The transfer, or supply and return between the two systems, is accomplished by the use of air operated flow diverting valves. These valves are operated remote-manually by means of a switch mounted on the ventilation panel in the Main Control Room. In warmer weather-,

the containment air coolers remain on chilled component cooling water supply unless a minor incident occurs. Malfunction Analysis A failure analysis of equipment and components is presented in Table 9.4-6

9.4.4-6 12-1-69 TABLE 9.4-6 CONSEQUENCE OF COMPONENT MALFUNCTIONS Comments and Components Malfunction Consequences

]. Component Pump casing The casing is designed for 220° F cooling water ruptures temperature; standard test pressure pumps is 200 psig and maximum test pres-sure is 300 psig. These conditions exceed those which could occur during any operating conditions. The casings are made from cast iron (ASTM A48); this metal has corrosion-erosion resistance and produces sound casings. Corrugated metal expansion joints are installed directly at the pump suctions and discharges. These joints isolate the pumps from forces and moments originating in the connected piping; in addition, the pumps are des:l,gned as Class I. Pumps are

missile protected and may be inspected at any time. Rupture by missiles is not considered credible. A relief valve is installed downstream of the automatic makeup valve for the system, so that makeup source pressure cannot be applied to the casings. All units can be isolated by valves and the standby pump can carry full load.

2. Component J.

Original pump Standby pump for that reactor unit cooling water fails to start can be used pumps

3. Component Standby pump Standby pump for other reactor unit cooling water fails to start can be started manually in Main Con-pumps trol Room, after manually reposi-tioning valves at the pumps.

4. Component Manual gate Prevented by pre-startup and operational cooling water valve at a pump checks. During normal operation each pumps suction closed pump is checked periodically, to-gether with its valves.

9.4.4-7; 12-1-69

TABLE 9.4-6 (Cont'd)

Connnents and Components Malfunction Consequences

s. Component Check valve at Valve is checked periodically during cooling water a pump discharge normal operation.

pumps sticks closed

6. Check valves , Stick closed One main is flowing at all times.

in supply mains The valves have split discs loaded at inlet pene- by light springs, and sticking trations closed is considered not credible.

7. Component Tube or shell Because of the system low operating

      *cooling wa.ter  ruptures              pressure and temperature, and heat exchangers                       Class I Design, rupture is considered unlikely. Each unit can be isob,ted

and carry full load. The standby unit intended for one reactor unit may be used for the other unit by repositioning valves. The exchangets, are protected from missiles.

8. Condensate Sticks open The automatic valve, located upstream::

makeup line from the check valve, is provided w:fitb check valve a handj ack; this permits manual clos'- ing of the. makeup valve, thus closing off the line. 9 *. Component Left open Prevented by pre-s-tartup and opera:t:fciria'1 cooling water. checks. On a unit in service thfs: heat exchanger c*onditfon would be observed by oJtet~ , vent or drain- atf~g personnel during routine ohser-' valve vation. On activation of a standby unit, the cond"ition would be observed by* personnel engaged in manually positioning valves at the exchanger

9.4.4-8 12-1-69 TABLE 9.4-7 RELIANCE OF INTERCONNECTED SYSTEMS Interconnected Purpose of Consequences if Inter-System Interconnection connection is Lose

1. Primary plant Clearing up plugged con- None - use of equipment is inter-auxiliary steam ditions on tube side of mittent. Also, it can be used waste disposal e*apora- in partly plugged condition.

tor bottoms cooler~ due to reduced temperature (component cooling water is on shell side)

2. Main condensate Makeup for surge tank The makeup line is double con-(in turbine nected to both main turbine room) generator units. Since it is unlikely that both units will be out of service simultaneously, the source is considered to have 100 percent availability.

3. 4. Boron Recovery Waste Disposal Signals to automatic con-trol valves in component cooling water subsystem piping Signals to automatic con-None - use of equipment is inter-mittent. None - use of equipment is inter-trol valves in component mittent. cooling water subsystem piping

5. Sampling Conduct sample to central None - samples at all important sampling station points may be collected at local sampling connections in the piping.

6.

Vent and Drain Disposal of equipment Since lines are open, without vents and piping drains valves or other devices, loss_ of the interconnections is not con-sidered credible.

7. Containment Signals to trip valves Valves fail safe (to closed Isolation. for isolation purposes, position) upon loss of signal under accident con-ditions

9.4.4-9 12-1-69

9.4.5 TESTS AND INSPECTIONS During the life of the station, the Component Cooling System is in continuous operation, and performance tests are not required. Standby pumps are rotated in service on a scheduled basis to obtain even wear. Following installation of spare parts or piping modifications, visual inspections are conducted to confirm normal operation of the system. Routine pre-startup inspections are performed along with periodic observation during operation.

9.4.6 MINIMUM OPERATING CONDITIONS Minimum operating conditions for the Component Cooling System are given under

Limiting Conditions for Operation in the Technical Specifications
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                                                                                                                              /    (FM-308,.

1"cc,51-151 l'.:cc-302... s1 ,. * -F cc-ao,-1s1

                                                                                            ~v mw,..,                                                                          GAS STRIPPER LIQUIC EFFLUENT                                                                                                                                                                              vos- ,oc-.~--~

1\*cc-2,i-1s1 W~STE GAS DIAPtlRAG.M t' VC:S-'09 COMPRESSORS

                                                                                                           *f-c.c--Z91-1!1                                                                                 5AMPLE ~OOL(Fl&

__Bll_. CCIOSA PRE55URIZER LIQUID r* SPACE

1t"!cc-299*151 .,* !!!

LIQUID WA5TE 1f-cc-!17-1s1 J(>--,,..-------...---~--1 t '" ....

                                                                                                                                                                                                                                                                                                                       ~
                                                                                                                                                                                                                                                                                                                       ~

1*-cc-270-1151 (l*BR-P-Y. eoa,.:>ttl E.~ D\Sf ~NP) CHill*P*;JB CCIOS A (l*BR*P.4 .. PRIIUJl'f DMtt, 'TA.NK PUII.\P) CCK>S B (l*BR-P-*B .* CC:.IOC." (1*811-P.C.A cc1at18 ( l*Bil*P.lio.

                                                                                                                                                                                                                                                                               ~EALWATEJI.                                                              CCIQ11'         CI-B~P."Tt.       &'TRIPKR C\QC PU~)

HlaAT E:XCHANCn.2 CCl:)!,8 CI-Blil*P-78 cc1oe"' Cl*LW-P*2., CCI088 C1-\..W*P*t.in CCIOII ll*Lll*P*4 -.slE l>l'iP ~V...P D1&1" PUMP) ccno (l*LW*P-e W&U 111aP ,v..p c111e PUM Pl in {l-l!IR*C*IA OrtllMEAD GoAS couPRH'IOII) Cl:BIK*IB J. 11 U U )

                                                                                                                                          ..,"                                                                                                                                                                                                                    DEiA.IL B i,
                                                                                                                                         .\)
                                                                                                                                             '                            1..*CC.*Z.7~*151
                                                                                                                                                                                                                              ~Ill
                                                                                                                                                                                                                              ,.,.FT!RCOOL!R HEAT REMOVAL fs-:,-!i-lC*N8
                                                                                                                                                                         ~**CC.*49*1'51                                             (FM 32A) 1
                                                                                                                                                                                                                     !. CC-300-151             f cc-29s,1s1 WF-12 ...

PI

                                                                                   . C.Cl53 SP,MPLE COOLERS 14°*CC-7t-l~I (F"1*2H)
    * *UD)                                                     VVF-IZA.--"                                                                                                                    19,*c.C*liA*IZI

- - - COMPONENT COOLING SYSTEM UNIT I - SHEET 3 SURRY POWER STATION

FIG. 9.4-4 OCT. 15, 1970 I' FS*--~L T[ 1s.*-cc-231-1z1 VVF*12A 1----~---=<~ J=. A 18"-C..:.-7-12.1 (FM-22C) CC ISSA (TYH) 24"-CC*2U-121 . ~ f < PI TOUflllTWO Z.

                                                                                                                                                                                                                                                                                            ----R'Tt> ~ -

TI CC.l<llA cc ,00 IB*-cc-,-,zi (FM*Z2.F) IB"*CC-8*1~1 (FM*22r,l PI T! k,&,c7c7'14Cro;;;e. c.c1sse TO UNIT NO I Pl 18"- cc- 10-121 (Fll*22c) t8"-c.l-'l28-IZ.1 \IVF.-12A TI 12"-CC-27-121 (FM*22c) 8"-CW-2/-ISI (FM-22C) IS"-C<'.-220-/Z/ 1_!--cr.-:l09-1s1 IIH~-2~*12 TI CC130A 6~CC-!20-l':il

                                                                                ~1s~,c.-z?a-rz..1   ~ 1s"-cc.-22s-111 r---,--~--~-D'f;J-,--~-~-----,-----t(F"'M-l',ZB)

Pl s*-c.w-ii--11:=>1 VOW-15A CCl9J (TYP) 1e*:..cc-G-IZI 1a"-c.c~2?i4*1'21

                                                                                                                                          .       PI               TI
                                                                                                                                            ~                  c::1ssr.

PA Pl L YGW-ISA COMPONENT coou11(, ',JAi EP HEAT EXCHANGER:, llJRl:llNE Re>OM 6"-CC-15'0*15'1 G"-CC-32-15"1

                                                                                                                                                                                                                 ~-.,.-,-.,-c---,.-----..,.---~--------:----s (FM*2ZF"J PI.                      PI                                                                                                                                     f - -......-~--°'"""""il (F"M*22F)                                        1t-cc-2s9-1s1
                                                                                                                              ~Cl49                 CCISO 1o*~cc-1sa-1s1 6'~C.C.- I 15 *ISi (F-B
                                                                                                                                                                                                                                &"*CC-118-151 6'-CC-IS7-\SI
                                                                                                                                                                                                                                    ~L
                                                                                                           '-vr:,s.eoos

__!:f._L_ ____ C-310-ISl CCl30B (FM*2'2C) 1rcC-30-\5.I s*-cc-1ic:Hs1 TO FLOOR f' CL-151 TO 6"*CC-14~1!il (FM*22F)""""\ DRAIN FLOOR DRAIN BORON PRIMARY DRAIN TANK VENT ~-CC-19'):ISI - I ', 1-CD-E.J-ZA 1-CD-EJ-ZB CHILLER CONDEN5ER A.UX eLD<; ff'!_

                                                ** ** CHILLED       Wl>."TER C.OIJUUJ:iER AIR EJ.ECTORS 6'-CC-117-151-e*~cw-1-1s1.

Ii"- CC- 2.G0-1 SI

~*

CHILLED COMPONENT COOLING PUMPS AUX. BLDG..

                                                                                                                                                                                                                                 -11 "-CC-2~0- ISi r   1J:*-cc-2ss-1s, l"-c.c;-24;)-151 Pl cciee'-

(."-CC-Zeto-151

                                                                                                                                                                                                  ~**cc-'2.87-151 PI                                        VG,S-'10!,~           ,,.--l\te:M-CC.-~'!.1-151
                                                                                                                                                                        ~                                                                        R:E"TUR~ i:'RCM I-BR-P-7AtB l*BR*P-4A4'B l*CD*P*2"                                                                                                                                                                                                                    1-LW*P*'lAlB l*BR-C-IA(B CHILLED WATER CIRC.PUMPS                                        REFUELING WATER STORAGE TANK COOLERS                                          CLISI 1                                                                                DET. B (FM*'l'2C.)

i'UR1311'1.E ROOt-'\ YA.RD ~ -ISr*CC.-14-12* (F'M-Z2C~-

                                                                                                                                                                                                                                 ,-1a*:.cc-1'5- 21              tff*1

2.2F; COMPONENT COOLING SYSTEM

                                                                                                                                                                                                                                  -!e"-cc-11:,- .. ,            (=-*1-22:A.,

UNIT 1.- SHEET 4. r*-c.c-24"!.-ISL SURRY POWER STATION

FURJ,.Jl5HED WITH CHlLLED WATER U"tilT

r FIG. 9:4-5 OCT. 15, 1970 3 22-~.:1-8171711 cc2oq I" 4*-cc-144-151 VGS-GO~ 1 PI 1:f-cc-131-rsJ l~ -CC-1'39*1SI ~- l'*CC-141-151 \

                                      *c~~,(

V60-<ooe.(TYP.) VGW-15~ l'-CC*l'27*1SI SEAL WATf-R HEAT EXCHA,,OEF\ 4'-CC-1-48-15'1 RV 10

                                                                                                                                                                                                       ~~---I-+     ----t TE-2144
                                                                                                                                                                                                                       . (11sqs*FM* I05B)

NON- REGENERATIVE HEAT EXCHAN~ER VCS-b013(TYP) 1'-CC.-38"l-151 l - . l - - - - - 4 ' - ( l l ! i 4 B *FM-2.1&) SAMPl.E COOLE'. RI; 14"-CC-72* 12 I s*-cc-144-1s1 C.LI E>I e*-cc-r4e-1s:1

                                                     \':.CC-13'4*151            l/i *CC-135-151                                   CLl'l..l CL\51      I
                                                                                                                                                                                                                    ~

PRt:SSURIZER LIQUID l"*CC-278-ISI SPACE

                                                                                                                                                                                                          \O~C-72-l'al

(.FM-22C) (,M-U.C) l'~CC-1'3?-ISI MAIN COOLANT COLC LtC: 11 14 -CC-14!)-1'21 GENERAL. NOTE~ ( REFERENCE DWGS F'M-22A

                                                  -r-cc-13Hs1 1*-cc-1,o-1s1 2{cc-121-1s1                    lf-CC*l35-ISI VCW-ISA C..Lll:,I     -CLIZ.I

)'!.CC-1'27-151 (FM-22C)

COMPONENT COOLING SYSTEM UNIT I. - SHEET 5. SURRY POWER STATION

FIG. 9.4-6 OCT. 15,1970

K

       .£Z-~.:l-8171711 rlJSIPE. IURBHJ,E
                                                                                                                                                                                                                                                                                 *uin=r uo. 2     '"Xl"'!

F'ROM CONT AlNMLNT ISOLATION. SYSTEM

I r;" *1 I i1

                                                                                                      ~   1154B - FM*22A
                      ~------;:

1 __1-1~548-FM-22. ----

                              ~~~/_        _j: it'~':".cl~~c /c~oc 1I                                     TV                   TV             I TV                                                                                                                                       l~S~*J51(1154P.-FM- IGA) mmi,l~                                                                 ,c~;o,c          CC205B             ,CC207            -l,cc2osA J;

r_ __/ ____ I_I r c-VE,M lSC.

                                                                                                                                                                                                                                                                                                 ~"** ~

_ _I , 1 ,1 __ I . !TYPJ

                     \ 14    \:      i  /2              J3   J     I             26            27 i         -      *err ---               25 - - - * - - -

HIJS-21 rM*ZIA

                                                                                                                                                                                                                                                                                           .  <TVe
                                                                         .-,,,_-o-,r*"-~--(\-"f'-~~~2f_...,___n__,,,..-,G,.
                                                                                                !ij    r* - -

6"-CC-181-JSJ I I t i"' I I H 8 I

                                                                                                                                                    ~H                  F
                                                                                                                    ~

F

t. I ccz.o~

VOW-161\----.,. I I Fl (TYi:'.) I

                                                                                                                  '-b           ,--            FT cczos C.C.2.'0'a sov cc:zoiii
  ~--r---                                                                                                                                       ~-VGW-l&A(TYP.)
   '(PM-*go}1;,-...r>tl!:"::::;;;~m:=::=:~~~-,                                                                                      ,,,o,-o,oc--
              \                                                                                                                 I !}J~ATAP-5 IFM-2aqt"'_.__ _ _ _....;;.;..__..:-_ _.._=...,..:...--;.--......-+,~--t-:--==-~-...,...---1:1ti---,..-..,..~                                                                                                                             2-CD*EJ-2A          2*C.D-EJ*2.8 CHIL:... ED WATER COr-JDENSER AIR EJECTORS FR.OM L _  CONTAINMEtH (PM-2iDM-----,

r:J- ISOLATION SYSTEM "le'*CC*IS-121 TI CDZC>Z IO'..CW*/4-151 6°*CC-1!18*1SI 1o*-cc-111-1s1 o,..J/ B' VIIW*l'St.(TYP)

                                                                                                                                                                                                                 ~/                  2*CD*P- 2B Q"-CC*l~~-151~                                                                                                                                                CD ao3&        CHILLED WATER I                                                                                                                 CIRCULATINB I                                                                                                                     PUMPS TO~'BlNE ROOM.

1-..........**-~~,:.c-1sa-,~1

                                                                           -+-+--+----+---                                                                                                                                                                                                                         GEN~RAL NOTES t Rt:F"ERENCE. DWGS.
                                                                           *.,o /       IL         9i I                         I I

(1154S*FM*'<?.Ze.) RTO CCZ22 TI CC222. TS CC222 TA H -* COMPONENT COOLING SYSTEM UNIT I -SHEET 6 SURRY POWER STATION

FIG.9:4-7 .

                                                                                                                                                                                                                                                                  . OCT. 15,1970 OUTSIDE REACTOR INSIO[ REACTOR:

CONTAINME~T COIIIT.h!~MENT RV CG!J9A VOS*'-OC (lYP.)

                                                                                                                                                                                                     ,*  VOS-WC (TYP.)                                              TO Ri:.-cl'OR f

CONTAINMENT

[TYP.) TO.UPPER BR6. LUSE OILCOOlCR (TVP.)- - _/ VGW*l5Pr 4 4 TO STATOR COOLER frYR)- - .../

                                                                                                                                                                                                    ,.                                                         RV
                                                                                                                                                            .." 1----i>()--J....-,

V u 3*

4 CC219B TO LOWEA 6RG. LUBE OIL COOLER(TYP.)- - -~-

t"

                                "                                                                                                                                                                   ,.                                                           ff
                                                                                                                                                                                                    +                                                                 VVf*l2A TO THERMAL          ,

BARRIE~ (TYP.)- - RES1DI.J+L HEAT VGS*40B(rYe> 1* ,.. VOS-IOOC(TYP.) REMOVAL EICHANGERS 1* voe-~ VGYrlSA

                                                                                                                                                                                                                                        .f._~_-J,,/--FFIUNNE.L l°,.

e 1'D FLOOR CRAIN 1' 1*

t:, LOCAL 5AMPLE DETAIL NOTES: i1i All TEMPERATURECONNE:CTIOHSTO BE !"EXCEPT AS NOTE~ NUMBER SHOWN AT REACTOR CONTAINMENT Lt NE

                         \!                   RV Pl                                                                                                                                                                     INDICATES RIPING PENETRATION NUMBER.
                         ~
                         *'      VGW-15A

~-CC-16'-151 I_" 1" Al.L P5i!.ES5URE, SAMPLE, Dl<A\N ~¥ENT cm.1w.ec.T\OfrJ'3 TO BS ~/4'WITH VO&-li,O C E.XCE.PT Af, NOTE.D, W FURN\5HED 8'< EQUIPMENT MANtlFACTIJRER t-lCB- Ml.IN COt,lTl

.: (TYP.)

I I~__ II ____ 2 ____ -__ 9 tfYP.)"'j ~ [FM-22A)

(TYP.)

," J: NEUTRON SHIELD TAI--IK COOLER5

81*21 B 1f-ws-s2-z1e SWIOIA 2**-ws 1.1 e.

2. WESTINGHOUSE COOLING -TER PIPING SUF'PI..IEC WITH CHARGING PUMP MUST BE MOOIFIEE) IN THE' FIELD 10 CDNNEC.T THE COOl..llolG WATCR UME!I CHARGIN4 PUMP AS SHOWN IN THIS FLOW Oll\GRAM.

3. CHILLED AND CONDENSER WA"lER FB-4tA'*

.ws-60*ZIB FI SWIOOI\

- \':-'.

9.5 FUEL PIT COOLING SYSTEM The Fuel Pit Cooling System .shown in Figur.e 9.5-1 pumps borated water from the spent fuel pit through heat exchangers and back to the pit to maintain fuel pit water temperature. Additional pumps are provided for purification through an ion exchanger and filter and for sur*face ,clarification.

1. Maintain the temperature of the fuel pit water below 140°F

when one-third of a core is placed in the *pit 150 hr after shutdown.

2. Maintain the fuel pit temperature below 170 F when one and two-thirds cores are placed in the *pit 150 hr after reactor shutdown. 'lb.is is the*worst heat load which can exist and decreases with decay time.

4. Maintain a minimum pool water level of 41 ft-2 in. which will provide a minimum water shield of 20 ft in depth (Section 11.3)

9.5-2 12-1-69 The spent fuel pit structure and associated systems are designed as a Class 1 system.

5-1.

TABLE 9.5-1 FUEL PIT COOLING SYSTEM COMPONENT DESIGN DATA Fuel Pit Coolers Number 2 Design duty, Btu/hr each 34,750,000 Shell Tube Fluid flowing Component cooling Fuel pit water water Design pressure, psig 150 100 Design temperature, °F 200 200 Operating temperature, max, °F 157 outlet 170 inlet
Operating pressure, psig Material Design Code 60 Carbon steel ASME III Class C 40 Stainless steel type 304 ASME III Class C

Purification Pumps

Pump casing Shaft Impeller Stainless steel type 316 SAE 4140 Stainless steel type 316

TABLE 9.5-1 (Continued)

5.2 DESCRIPTION

The Fuel Pit Cooling System has two full size shell and tube coolers and two full size circulating pumps and two full size purification pumps, all located in the fuel building. The coolers and pumps are arranged for cross-connected operation, if necessary. The coolers are cooled with component cooling water. The purification pumps take suction at the outlet of the fuel pit coolers and pump water*to an ion exchanger and filter located in the auxiliary building. The ion exchanger or the filter can be bypassed if not required. The water returns to the fuel pit at the far end opposite the suction point to assure mixing. The purification system is run independently of the cooling system whenever purification is required. The surface of the water is kept clear of floating matter by two skillDllers connected to two skimmer pumps. The pumps dis-charge to the skimmer filters, after which the water returns to the far end of the pool. The lowest level of pipe penetration through the fuel pit structure is 20 ft above the top of stored fuel elements. 9.5.2.1 Components All piping, valves, and components of this system which come in contact with the fuel pit water are austenitic stainless steel.

9.5-9 12-1-69

9.5.3 DESIGN EVALUATION 9.5.3.1 Availability and Reliability Two full size circulating pumps and two full size fuel pit coolers are provided

 .to assure 100 percent backup for cooling requirements. Redundant piping is provided from the fuel pit through the pumps and coolers to the main return header located above pool water level.

9.5.3.2 Purification of Water The 150 gpm filtering rate of the purification system results in a refueling ~ water cleanup half life of 2 days and maintains suspended solids at a low concentration for optical clarity. The skimmer filter removes particles which fall and float on the water surface. This reduces the amount of impurities which enter the water and also reduces surface refraction. 9.5.3.3 Storage Pit Water Leakage Control Slow leakage of water from any point in the piping or components or the cooling or purification .systems can be stopped by valves mounted close to the pool pene-trations. An alarm is provided on the pool to sound at a level loss of approxi-mately 1/2 ft; .this provides ample time to isolate the leaking equipment. Further, a large piping system leak can reduce the water level in.the pool to only 4 ft below normal since at this elevation the water level is below the pipe penetrations in the pool wall. This minimum water level ensures at least 20 ft of water over stored fuel and provides ample shielding and cooling.

9.5-10 12-1-69 9.5.3.4 Malfunction Analysis

Component Malfunction Comments and Consequences Spent Fuel Pit Pumps Pump fails to start The standby pump will be started manually. Circulation will be in effect prior to fuel being placed in the pit. If the operating pump should stop, over 1 hr exists to start the standby pump before the pool heat.s up 10° F.

Fuel Pit Coolers Loss of function The standby exchanger will be used. More than 1 hr exists to realign the piping system because of the slow heatup rate of the pool. The rea-lignment is ~ffected by opera-ting manual valves. Pumps, Coolers, Leaks of any size A slow leak (less than 100 gpm) Piping, Valves and will permit over 2 hr to other components isolate the leak before loss of 1 ft of water. A large leak can only reduce water to the lowest pool penetration which is at a high enough level to assure adequate shielding.

9.5-11 12-1-69 - 9.5.4 TESTS AND INSPECTION The fuel pit level and temperature instrumentation are calibrated on a periodic basis. Periodic visual inspections and preventive maintenance are conducted on all system components. Periodic sampling of fuel pit water is conducted

FIG.9.5-1

                                            ..                                                                         "                                                    J                           IC OCT. l!),1!;170 2;!*FP*13*152 rueL PIT COOLE'FiS VBS-15'1 l'l."*JP*4*15l.,

1'11>>\*f>>\.,5-z) (FM*oOOA) 2{-FP-13-152

                                                                                                                                                                                                            \/D5*15Y HO,e: CONfrlE'CTI ON A.C. WITH ILANK FL~N'OE 11.**0*lHSlt.

SPENT FUEL PIT 1r.*-i,,-1e-111 PUR\F"ICA.TION PUMPS 1 t**PG-.2.9*1GI (FM*29C)

                                                                                                                                    ~-- -2.'-FP-2.2.*112.

2'*FP*21-152 R£f'llll:t,ICC CWGS I I

                                                                                                                                                                                         . COMPOHE'IT COOUNC 'NATtP. s~mM BOQ.ON ,-,c.COVcM~ STSTlM nuac Jll*2~C wm< DISl'O'!AL. SYSTEM                    Plll-90C VCNT4 DR,.llt7 ~ffTEM                     f'M-IIX>>.

NOTCS*

                                                                                                                                                                                           ~YMIIOL A,C, DENOTES .ADMINISTRAT\11£ (:DNTROL VAlVE WEIR SPENT FUEL. PIT 2'1'1'*2.0-15'2.

IOIA

                                                                                 '---------            . --'---*                       E>!;ll"'D-~--Jnsc".80)
                                                                                                                                                              £V8HSY f /sc~.e.o>

i N

                                                                 .t ..
                                                                                                                                                          ~*csc:~~~-.,... ,n
                                                                     'I' Pl                   ~

FC\0.3A

SPENT rurL PIT SK~t,J.ER SPENT f'l/E'L PIT ~IUMMER ,PU...PS FILTER SPENT F'UtL PIT PUMPS 2*-aa.-s11-111.

(FM*IOO") FUEL PIT I~ 2~ DA*"* 152 (FM*IOO...) COOLING SYSTEM SURRY POWER STATION

9.6.1-1 10-15-70 9.6 SAMPLING SYSTEM The Sampling System as shown in Figs. 9.6-1, -2, -3 and -4, is designed to provide primary and secondary fluid and gaseous samples for laboratory analysis, and continuous radiation monitoring of the steam generator blowdown. Con-ductivity,pH,and dissolved oxygen are also monitored at diffel;'ent stages of the I I secondary cycle. 9.6.1 DESIGN BASIS Process fluids and gases are representatively sampled for testing to obtain data from which performance of the station, equipment and systems may be determined

Samples of process fluids and gases associated with both the primary ~d secondary system are either taken periodically or are continuously monitored!

Two general types of samples are obtained by the Sampling System: high tempera-ture*samples (> 150° F) such as th~ Reactor Go~lant System samples, and low temperature samples (- 150° F) as from the high level waste drain tank. The various samples taken are listed in Table 9.6.1-1

9.li.1-2 12-1-69 TABLE 9.6.1-1 SAMPLING SYSTEM SAMPLES High Temperature Samples From Each Unit

l. Pressur:Lzer vapor

2. Pressurizer liquid

3. Residual heat removal liquid taken downstream of the residual heat removal pumps

4. Residual heat removal liquid taken downstream of residual heat exchangers

5. Hot leg primary coolant taken from ea~h of the reactor 6.

coolant loops Cold leg primary coolant taken from each of the reactor coolant loops

7. Steam generator hlowdown liquid taken from each of the blowdown lines

8. Flash evaporator pump discharge

9. Flash evaporator distillate

10. Main steam taken from each of the main steam lines

11. Steam generator feedwater

                                                                    ,9 .:6.1-3 1:2-:1-;6'9

TABLE 9.6.1-1 (Continued)

High Temperature Samples Common to Both Units

1. Auxiliary heating deaerator

2. Auxiliary heating boiler lower drum

3. Auxiliary heating boiler steam drum

4. Liquid waste ev.aporator Low Temperature Samples from Each Unit

1. Supply header to Chemical and Volume Control System demineralizers

2. Chemical and Volume Control ,System cation demineralizer effluent

3. Condensate ~wnp discharge header

4. Chemical and Volume Control System dehorating demineralizers effluent

5. Chemical and Volume Control System mixed bed demineralizer, effluent

6. Volume control tank liquid

7. Volume control tank gas space

8. Pressurizer relief tank gas space

9. Flash evaporator demineralizer effluent

9.6.1-4 12-1-69 TABLE 9.6.1-1 (Continued) Low Temperature Samples Common to Both Units . 1. Low level waste drain tanks liquid

2. Boron Recovery System test tanks liquid

3. High level waste drain tanks liquid

4. Boron recovery tanks liquid

5. Component cooling water

6. Primary drain tank liquid

7. Gas stripper liquid effluent

8. Primary water tanks 9.

10. 11. Contaminated drains collection tanks Waste disposal evaporator test tanks Gas stripper surge tank gas

9.6.2-1 12-1-69

9.

6.2 DESCRIPTION

All the sample lines coming from within the containment contain high tempera~- ture samples with the exception of the pressurizer relief tank sample. Where two or more samples join into a common header (i.e~, the primary coolant cold leg samples), each individual sampling line has an air operated valve in the ! line which can be remotely operated from a control board in the auxiliary building sampling room. The primary coolant hot leg and cold leg samples flow through delay coils prior to penetrating the containment. These delay coils permit sufficient decay of nitrogen-16 so that these samples can be handled in the sampling room. All sample lines penetrating the containment, with the exception of the residual heat removal sample line have two automatically operated valves in the line, one Just. inside and one just outside the containment. The residual heat removal sampling* *line has just one trip valve, located outside the containment. These trip valves close on a high containment pressure signa"l. The high temperature samples pass through sample coolers located in the auxiliary building sampling room. These coolers cool the high temperature samples to a: temperature low enough fo-r safe handling. Sample flows- leaving the cooler are manually throttled and can be directed to a purge line or to the sampling sink. The pressurizer vapor space samples in addition pass through capillary tubes which limit th~ flow of steam. The sampling lines from sampling points outside the containment but inside the

auxiliary building also discharge to the auxiliary building sampling sink with

9.6.2-2 12-1-69 the exception of the liquid waste evaporator sample. This sample line is located near the evaporator as it requires heat tracing. Sample lines from sampling points in the turbine building discharge to one of the turbine build-ing sampling sinks (one for each unit). The high temperature samples also pass through sample coolers and are manually throttled. In general, samples can either be directed to a purge line or to the sampling sink. The main steam samples also pass through capillary tubes. The purge flows _of the various samples are discharged to the volume control tank, the Vent and Drain System, or elsewhere as appropriate. The radioactive samples in the auxiliary building sampling room discharge into hooded sampling sinks. In addition to the above facilities for periodic sampling, there are facilities for continuous radiation, pH and conductivity monitoring of the steam generator blowdown samples and oxygen, pH and conductivity monitoring of the condensate. pump discharge, and pH and conductivity monitoring of the feedwater. Radiation monitors in the steam generator blowdown sample line detect primary-to-secondary side leaks in the steam generators. Monitoring of the condensate pump dis-charge is required for detecting tube leaks in the condensers.

9.6.3-1 12-1-69 9.6.3 DESIGN EVALUATION If a critical sampling line becomes inoperable due to some malfunction, there is at least one alternate path which can be used to obtain a similar periodic sample or for continuous monitoring. For example, if the condensate pump dis-charge sample line becomes inoperative, condensate can be monitored con-tinuously for conductivity with the local condensate hot well sampling lines. If one of the steam generator blowdown radiation monitors malfunctions, a second similar radiation monitor in each unit can be used. If one of the steam generator blowdown sampling lines becomes inoperative, the condenser air ejector radiation monitor provides indication of a steam generator primary-to-secondary side leak

9.6.4-1 12-1-69
9.6.4 TESTS AND INSPECTIONS Most components are used regularly during power operation, cooldown and/or shutdown, thus providing assurance of the availability and performance of the system. The continuous monitors are periodically tested, calibrated and checked to assure proper instrument response and operation of alarm functions
FIG.9.6-~

    ..                       n                           C            I            D            I            E             I            F             I            .                                            H                       T                                                                                                                    K                            I OCT.1~ 1 1970 CLJ-~~51.15*1-,1~()'3 I

r-------------------------------------------------------------.. . ------------------"------:,::-.-------Q.-<:::-i---<l"<1-1-l.!A*l::ll. t>UillNf.RAL1Zt.li:S .w"F1...UEN"T sA.MP .. E *fs<:>-1i-1-c.-Ne. "I-I.EM Wl..cauT ~.,.s IJl\lCED BE"..'i) DC.l.A.l.lE~jt,,.LlZ.E":cS ~F"'C'._,JEl,..,;"T ~M?,...£5

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9.7.1-1 12-1-69

9.7 VENT AND DRAIN SYSTEM

.The Vent and Drain System collects potentially radioactive fluids and gases from various systems and discharges them either to the Waste Disposal System (Section 11. 2) or to the Boron Recovery System (Section 9. 2).

DE~IGN BASIS The Vent and Drain System is shown in Figures 9.7-1 thru 9.7-4. The *drains are separated into those carrying waste fluids to the waste drain tanks for pro-icessing and disposal, and those. carrying reactor coolant .fluids to the primary

*drain transfer tank and primary drain tank for p~ocessing and recovery. The

vents are separated into vents in which air is the predominant gas (filtered and discharged to the atmosphere), and vents in which hydrogen and radioactive gases are the predominant gases (discharged to the Gaseous Waste Disposal System).

Redundancy has been provided for all active system component~ to ensure sy~tem operation. The Design data for the Vent and Drain System Components is given in Table 9.7-1. The primary drain transfer tanks, primary drain coolers, relief valves and the piping, valves and supports of the Vent and Drain System confonn to seismic

£lass I criteria

9.7.1-2 12-1-69 TABLE 9.7-1 VENT AND DRAIN SYSTEM COMPONENT DESIGN DATA Primary Drain Transfer Tanks Number 2 (one for each unit)

Capacity, gal 600 Design Pressure, psig 200 Design Temperature, °F 400 Operating Pressure Atmospheric Operating Temperature, °F 150 Base Metal Material ASTM A515 Gr 60 Clad.Ung

Stainless Steel Type 304L Design Code ASTM III Class C Primary Vent Pots Number 2 (one for each unit)

Capacity, gal 20 Design Pressure, psig 25 Design Temperature, °F 200 Operating Pressure Atmospheric Operating Temperature, °F 200 Base Metal Material Stainless Steel Type 304 Design Code ASTM III Class C
9.7.1-3

. 12-1-69

Valve Pit Leak-Off Pots Number TABLE 9. 7-1 (Continued) 2 (one for each unit)

Capacity, gal 5 Design Pressure, psig 100 Design Temperature, °F 250 Operating' Pressure Atmospheric Operatinc Temperature, °F 105 Base Metal Material Stainless Steel Type 304 Design Code None High Level Waste Drain Filter

Number Retention Size, microns Filter Element 1

5 Fiber Capacity*Normal, gpm at 2.5 psi ~p 50 Capacity Maximum, gpm at 5 psi ~P 75 Material Stainless Steel Type 304 Design Pressure, psig 150 0 Design Temperature, F 250 Design Code ASME III Class C
9.7.1-4 12-1-69 TABLE 9.7-1 (Continued) Low Le:vel Waste Drain Filter Number l Retention Size, microns 5 Filter Element Fiber Capacity Normal, gpm at 2.5 psi ~P so Capacity Maximum,gpm at 5 psi ~p 75 Material Stainless Steel Type 304 Design Pressure, psig 150 Design Temperature, °F 250 Design Code ASME III Class C Safeguards Area Sump Pumps Number 4 - two for each unit (1 required) Type Vertical Centrifugal Single Stage Motor Horsepower, hp l Seal Packing* Capacity each, gpm 25 Head at Ra~ed Capacity, ft 39 Design Pressure, psig 150 Design Temperature, °F 180 Materials Pump Casing . Cast Iron Shaft Steel Impeller Bronze
9.7.1-5 12-1:-69

Fuel Building Sump Pump Number TABLE 9. 7-1 (Continued) 2 (one required)

Type Vertical Centrifugal Single Stag~* Motor Horsepower Rating, hp 3 Seal Packing Capacity , '* gpm 25 Head at Rated Capacity, ft Design Pressure, psig. 150 0 Design Temperature, F 350 Materials Pump Casing Stainless Steel Type 304 Shaft Stainless Steel Type 30.4 Impeller Stainless s*teel Type 304 Auxiliary B.uild:i.ng Sump Pump Number 2 (one required)

*Type Vertical Centrifugal Single Stage Motor Horsepower, hp 2 Seal Packing Capacity, gpm 50 Head at Rated Capacity, ft 49 Design Pressure, psig 150 0

Design Temperature, F 350 Materials Pump Casing Stainless Steel Type 304 Shaft Stainless Steel Type 304

Impeller Stainless Steel Type 304

9.7.1-6 12-1-69 Number TABLE 9.7-1 (Continued) Reactor Containment Sump Pumps 4 - two for each unit (1 required) Type Vertical Centrifugal Four Stages Motor Horsepower, hp 3 . Seal Packing Capacity, gpm 25 -Head at Rated Capacity, ft 80 Design Pressure, psig 175 0 Design Temperature, F 225 Materials Pump Casing Stainless Steel Type 304 Shaft Stainless Steel Type 304 Impeller Stainless Steel Type 304 Incore Instrumentation Room Sump Pumps Number 2 - one for each unit Type Vertical Centrifugal Single Stage Motor Horsepower, hp 1.5 Seal Packing Capacity, gpm 10 Head at Rated Capacity, ft 40 Design Pressure, p~ig 150 0 De3ign Temperature, F 350 Mat:erials Pump Casing Stainless Steel Type 304 Shaft Stainless Steel Type 304 Impeller Stainless Steel Type 304
9.7.1-7 12-.1-69

TABLE 9.7-1 (Continued)

Component Cooling Heat Exchanger Pit Sump Pump Number 1 Type Vertical Centrifugal Single Stage Motor Horsepower, hp 1 Seal Packing Capacity, gpm 25 Head at Rated Capacity, ft 44 Design Pressure, psig 160 0 Design Temperature, F 180 Materials Pump Casing Cast Iron Shaft Steel Impeller Bronze Primary Drains Transfer Pumps Number 4 - two for each unit (1 required) Type Canned Horizontal Centrifugal Motor Horsepower, hp 3 Seal Canned Pump Capacity, gpm 60 Head at Rated Capacity, ft 64 Design Pressure, psig 150 0 Design Temperature, F 400 Materials Pump Casing Stainless Steel Type 316 Shaft Stainless Steel Type 316 Impeller Stainless Steel Type 316
9.7.1-8 12-1-69 TABLE 9.7-1 (Continued) Loop Drain Header Relief Valve and Primary Drain Transfer Tank Relief Valve Number 4 - two for each unit Capacity (each) gpm at 150 psig, 366° F 100 Pressure Setting, psig 150 Design Pressure, psig 150 0 Design Temperature, F 366 Primary Drain Cooler Number 2 (one for each unit) Total Duty 5 x 10 6 Btu/hr Shell Tube Design Pressure, psig 150 200 Design Temperature, °F 200 400 Operating Pressure, psig 100 50 Operating Temperature, In/Out, °F 105/140 350/150 Material Carbon Steel Stainless Steel Type 304 Fluid Component Reactor Coolant Cooling Water System Drains Design Code ASME III ASME III Class C Class C
9.7.1-9 12-1-69

TABLE 9.7-1 (Continued)

Vent and Drain Piping and Valves Materials Stainless Steel and Carbon Steel Design Code ASNI B31. l Design Pressure, psig 95 0 Design Temperature, F 250

9. 7. 2-1 12-1-69

9.7.2 DESCRI*PTION Liquids with high levels of radioactivity, other than those orlgi.nating in the Reactor Coolant System (Section 4),are gathered and transferred to the high level waste drain tanks in the Liquid Waste Disposal System (Section 11.2.3) by the high level waste drain header. Liquids with gross activity levels of

-3 ' '

1. 5 x 10 µc/ml or greater are classified as high level.

3 Liquids with low levels, i.e., less than 1.5 x 10- µc/ml of radioactivity, are transferred td the low level waste drain tanks in the Liquid Waste Disposal System by the low level waste drain header. Both containment structures, auxiliary building, fuel building, both safeguards areas, the component coolers in turbine building, and both incore instrumenta-tion areas have been provided with sumps for collecting drainage. The drainage is transferred by sump pumps to either the high level or low waste drain tank, depending on the activity level. If the activity level of the drainage is unknown, it is transferred to the low level waste drain tank for testing. If the test reveals a high activity level, the drainage is transferred to the high level waste drain tanks. The containment sump collects all drains in the containment which do not originate from the Reactor Coolant System. The auxiliary building sump collects floor drains, *equipment drains, ion exchanger drains, and filter drains. The fuel building sump, safeguards area sumps and component cooling heat exchanger

pit sump collect floor drains in the respective areas

9.7.2-2 12-1-69 Drain liquids originating from each Reactor Coolant System, are discharged to a primary drain transfer tank through a high pressure drain header. The high pressure drain header permits high or low pressure gravity draining of individual reactor coolant loops, the pressurizer relief tank, or the complete Reactor Coolant System, except for the reactor vessel. An alternate use of the high pressure drain header is to provide a path.for draining the loops during hot shutdown.

Low pressure radioactive drains, pressurizer relief tank drains, and leakoff liquids from valve stems and reactor coolant pumps drain by gravity to the high pressure drain header through the primary drain cooler to the primary drain transfer tank. From there, they are pumped to the primary drain tank in the Boron Recovery System (Section 9.2) by the primary drain transfer pumps

The primary drain cooler is provided to cool all liquid entering the primary drain transfer tank. A high temperature alarm is provided in the primary side of the cooler outlet to warn the operator of excessive hot liquid flowing into the primary drain transfer tank.

The sample header drains flow directly to the primary drain tank. In the event of high level in the volume control tank of the Chemical and Volume Control System (Section 9.1), the demineralized letdown flow is diverted directly to the primary drain tank through the primary drain transfer pump discharge header. An*air vent header is provided in each reactor containment to vent the Reactor Coolant System and components during filling operations. A vent pot located at the end of this header separates any entrained liquid for drainage by gravity
9.7.2-3 12-1-69

to the containment sump. Air leaving the vent pot is discharged to the Gaseous Waste Disposal System (Section 11.2.5). Vents from the ion exchangers and demineralizers, the component cooling surge tank, and waste drain tanks are handled in the same manner.

Radioactive gases are vented to the Gaseous Waste Disposal System. Included in this category are vents from the pressurizer relief tanks, volume control tanks, reactor coolant pumps standpipe vent, bypass vents, and the Sampling System gas sample purge line. Piping for the _Vent and Drain Systems is designed in accordance with ANSI B31.1, Code for Pressure Piping. Isolation valves are provided in all

vent and drain lines from the containment structures; for details see Section 5.2

9 .. 7_.J-.1 12-1-69 9.7.3 DESIGN EVALUATION The Vent and Drain System is sized to handle the maximum amounts of liquids and gases expected during station operation. Sizing the equipment for these maximum values results in design parameters shown in Table 9*. 7-1.

Austenitic stainless steel piping is used to transfer liquids and radioactive gaseous waste; carbon steel piping is used for nonradioactive gases. 9.7.3.l System Reliability The auxiliary building and fuel building sump pumps are a duplex pump arrange-

ment. The pumps, which are full size, are controlled by float switches which cycle the pumps off and on.

the pumps. An alternator is provided to obtain equal wear on Two additional float switches are provided; the first one starts the standby pump in the event the operating pump fails and the second float switch sounds an alarm on high sump level.* The containment sump pumps are a duplex pump arrangement. Each pump is full sized and independently controlled. One pump is in automatic service and the other in standby. When the water level in the sump reaches a specified height, an alarm sounds and the pump starts. The pump stops automatically upon emptying the sump. Containment isolation valves are provided in the discharge piping and are interlocked with the pump controllers. The isolation valves open and close on pump start and stop. When initiated, the containment isolation signal overrides the pump start signal to keep the isolation valves closed.

9. 7. 3-2 10-15-70 The primary drain transfer pumps are full sized and independently controlled.

Two pumps are provided for each unit. One pump is in automatic service and the other on standby. When the water level in the tank reaches a specified height, an alarm sounds and the pump starts. The pump stops automatically upon emptying the primary drain transfer tank. Containment isolation valves are provided in the discharge piping and are .interlocked with the punp con-trollers. The isolation valves open and close on pump start and stop. When initiated, the containment isolation signal overrides the pump start signal to keep the isolation valves closed. The primary drain coolers and primary drain transfer tanks and interconnecting piping, valves and supports are designed as Class I components for earthquake protection. They are also protected from the design tornado by being located inside the containment structures~
9.7.4-1, 12-1-69

9.7.4 TEST AND INSPECTION Formal testing of this system is unnecessary since it is in normal day-to-day operation. Inspection is performed in accordance with normal plant main-tenance procedures

FIG. 9.7-1

.. 10 11 'l" FEB. 11, 1972 \1'001-IN..:!-8t71711 VNITI-Jo,1 1>>s1oe l'Of.ffAINME.WT FUE.L. PIT I ON t.,)(C PRIMARY DRAIN (FM*31A TANK 1-&1;:-TI<-!.

(FM*!!A) CESIU~ l=tEMOV ION EX.CH FM-29

.5EAL WATER l"-1'-'FILT~

CESIUM RC.MCW VSNT ( FM*29B) ION EXCH (FM-29 3":DG-41-152 1"-VA-1115-\54

,t '",<N . i 0-\/A IS4 1 - - - . - - - - f ( F M

308)

Jp*VA*32*154

~-

COMP COOLING SURGE. Tl< (FM*3DA) r-VA-34-r s_*_..__ _ _ _ _..,1.,.____; 0 Ill (FM-220) I n-tsA c L-111.ir f!O Q l~ 4DG-2'4\-IS2 FROM VN\T 2i e"-VA-45-154 vt>S-IS'Z if

4."' z 0

..2, (11541!1-FM*IOOAl~ )

VDS-1!..Z 3_~ ~ '-......:

---- ----........... ,...._....___.___....____._o----; *&t:~~;:;

0

~" ~ ., 0 uJ I-'-. 5-f.fu-t!=!\Jl-32/\) --,-----------------1-------'------*1-----!( FM-1071>)

r?i~\/A.-1.1s4

"~ a'" "'o "'~ ~)I 2."*VA-11'5*1S4-FROVi u..i1T C..

0 11s.. a- fM-100.-.. 1-0G- 30-152 c~1s1-..i--z:L15'2 l"-CC-293-ISI I ,~-OA-91 15'Z. t'<l\*&9*1~2 J:' .... ,;_05 ,l"-DA 1-152 (FM*ttG) I

Z"*DA-(.*ISZ.

Z"*DA*'-*15'-____.,/ HIGl-l Lfll'-LSAMPLC. (FM*I05A) PUR~! MP2. 2'-FP.*Jt:.,*/52 (F"'1-32A,)

),-~----.----""='-'-'"'-""-"--< LOW 1-EVEl. SAMPLE PV'l=l:G.E. Ht>R, r-- ----- - -*-, {FM-U/\\

AlON'\ c'ONTAINMENT

~LA"TION SYSTEM wrn-\ PUMP 11.JTElc!UXK r;; l"-()6.*SS-152 NO~E.13:_,,

J 2"-DA-121-11;2 (115.tl8-F'M-100A) LMC*! CAPPED LEAl<ACE MOITTTORING CONN

~FM-1<>7B\~ VD~*ISY- )I{ 0E.NOT~'5 FE:NE.TRA..TION NO. (TYP)

('TYP) ALL PRESSURE TAPS (P:I 1 PT,PDI,PCV)TO&Ef*

,----------------------------t:l/:J-4--,--C-\-.....J--------------..;..*--*::J-S.~-!::D-:!;A.:.*i88i:--1~s]2C::;::-_ *

DE.tJOIE.S '.5UP~IE.O* wn.+.PVMIPS .

"-'*OA-26'-ISt!:

REFERENCE DWGS:

",'t><.*l"*IH VENT f DA.I.II{ SYSTEM '(FM-30.A.) ,__,*_-_D_A~*SO_*~JS~'-_ __. EM-1001' <FM-1011')

COMPONENT. cooUNG WATER r::::-:** I BORON RECOYeFtY. Sl'ST'EM FM'22A,Uli rM-2U.,*4c

.~~~=-:.: '.*:

WASTE DISPOSAL.. SVSTEM FM-30A,1'14C. F"U(.L PIT SYSTt:.ht *FM"'l,IA SA.MPLE. SYSTEM FM-32A 1i>z:o.-eo-r,z REAtTt'IR Ceol-A.NT S°f'ST&"C . FM-103AU.-

'l *0.*11,1-1$2 RESIDUAL HE.AT REMov.a.l. .F.M-104A '

cs=====:: lj':.O(o-53*1

. CHEMIC,!t,,L ~ .VOI....UME. CONT SAFETY IN<J<:CTION SVSTE_M* *svs. .Fl.1c10'>A,8~C .FM.:.106A = t*OG-ec.-"e GASEOUS WASTE Dl~PCJSAL ft.1-107,._

C.Hf.M F'E.<:D , STM G~N .IILONDOIIIN fM-]22. W

~

C -~**°'-*91-1SZ

't'\P111,,*1>t*IH 1*-DA*S--1-152.

Rt:ACTOR C"'VITY PUFllFICAi°"IOl't SYS. PENETRATIONS FM-114A. FV*\A VENT ~ORA<N* SYSTEM IIS-48-F"1*100,\(I:> 2**0A-ce-1se l"M*IOI/\ '.1z.*C<.*9',*IS'2 :1 11 -DA-,1~1$.2. e CHEMICAi. 'IOl...l,1"'1E CCWT SYS. 1154~-FM.:.tOs... (&

'2"-0G.*97*1,z CFM*:ffl'<l llz~CS-~*1!12 ~-oc.-90-isz VBS-ISY ~

I"- -152

.~r-DA-21-ise ~ ,:. < 0 <'I

l'-DA-49-ISZ.

"~ :i"' ,_ -b *'

1--~-='-"-"-'=-.-..J

"~ 1"-~*SO-ISZ m,i, t,s, H " 1--.....:..!C:,.::,c:c..=:...-_,,,-

l'*DA*'-4*1!5Z -C,M*30~)

==....... --r'---.--"-.---------- ~ ~ PIT O~AIN5 N

m

!:QUIP. Oil. l.l NE:: !F&,9A) m $ ~

(I g ~ a;

~

l*D.O...*P-'3A¢ 35

~ g. g g VENT & DRAIN SYSTEM f-06,.-P-'-

COV.POt.JEf,,JT CO:,L.1 i;.,:s H.T E:XCH. PIT 'SUMP PP

~-[;;¥\..- ~- '2.A PI...IEL ~IU)lt,.JG ~UMP ot .,2.e*

PUMP::S AUX 5L>MP 8ULL0\N(a, PUMPS ~{t ; Cf' *

.its l.;. ~ ;:po l>>JIT~.I W:51DE t)t,JIT l,JO.i (FM*I05B) (11548-FM*JO*e) UNIT SHEET I OUT~IDE COt.JTAlt-J. U",IIT l<JG. I 2 SURRY POWER STATION DOEO LINE.S f CH GD - DESCIHl"JION DUCRll'TICH CL ~SS\FICATION UCtU.'110

H IC SF

FIG 9.7-2

'800J-V\J.:l-8v v 11 0C115, 1970

2.':0G-l?*ln 2'~0(,. 1'2.*IS'2.

LO VDS*\SY 1*-~1-G..:..* JS'Z {Fl'l'\*1068) LO V~S*ISV

,--11-----------'-C __ z:_-o_<o_*3_7-_1_s2__ ,._____,_.-.'!.1.?1i1-1s-,st (fM.*104A) ~~CU, t'l4-/1;,2 ,,.,DG*\*152

{,::t.t, 1oi;.c) CL*l!>O!. CL,1St r>{>"'*RC*l:1so2. DE.T.A.ILA (TYi') ,.

~-VG*8*1SZ. ( *~-lO~A) LOOP CR1'..lt>J \-\'OR I~ 2~0G-\B*1Sa i~*RC*l:150~ fVG*9*152 VALVE STE.tJ\ \.E.~\\On,,; VDS-15Y I~ ~ ~ N ~

N

-tYf.'::DG*41-IS1.

ol

~ "' IJ>ll"t."I lN:,10\i-LMC 0~* RC.~*1502 3*

4 *VG *10 *IS2 N

~ !:: !!l m ! ~ "' "',;,J ~ ,i (FM-22~)

4C.C-113*1SI*, c.o>IT A, I tJ ME.IJ T

~ ~ .(

I~ !!)

'u" ~

I

-"~

D

,,;,0 ,,.""'

0

~

0 J'"

;jj g

j;

~ ~ 0 ~ ,..:,?

I

~

(FM-22.:) H 4'*CC*\05*15\ L"

~I01tfl ..~ _..J "" 0fRc*;:*1so2 !? /~ f*VG*ll*\52. ~ "" ~

DC.>010 I

,. 'j p .. . ... ~- ~ ~

r-LI De. 101 I I 04*RC~~~-1502 4 -'IG-W.*l!ra. *' :;, 0: I~ "' ___J ,*-----j 0~~07 :

~l.RC*l43*IS'02.

.-ve.-1,-1u FM*IO~&

VALVE 5Tt.lA LE'-\1.0l'Vi 5\ t:81 lr--1 "' !l!

"' ~ ~

1"- 5I

I05

l5Z i ,..,~ r- .,.0 pR:1µ.AR.'.'( .Di-1~\N

.....~ 'T.~~Sl'EA. "!"'-"... TV (FM- IO~e.) ,." 0 ~,*,.8 .\l " foc-1n-1*s2 DGIOBA ~

mov C,u DETA.,\L A ,,

.1n--1 I<Tyi f I s"* R~ao-,soz 1 3 .....~'"*..-DG--*&5-__ vcs 1,011 Ii' ,.,;. f'DG*/Zl;"l lTYP) FM*

103B{ . f-Du*ll!.* 4 -DG-~*-1~2 ~ t>l:i t>l:i Ji t*l I f*OG*lt>*lll

---------~ ~DHSY(TVP)

HHO~A FM-10~ B l*OC.*P* 11'.\IB F'RIMAP-Y DRAIN UNIT"'t lNS\OE fDG*<,,7-152 . L-.._j *'°"_f RC.*114*1500 54 t---..----,.~*- r--,{ _;_ . ..., I.. -,,

~ ,rvA-10-1~4 2-"*V~*l"3*154 TRAt,.ISF'E.R .. PtJM.?S CONTAINM~NT 2:"*VA.-l'S. IS I 5 -DU-b6-IS2 t-----i' 9*'

s~-RC.~W-1502 ~ "'

;,- ND,E~:

GrnERAL 1-lOTEs, / REFERE.['ICE OW(,,~ FM*ICOI< Nr~INMENT NOTE"A" ~POC'.. PlttE. *10 SE. U";:.E.t> FCA HI S-i\...-Rc.:.:,se-1so2 DRA\t...1\t.J.~ W\-\E.N R~iO~ \~ 'SH1.Y"TCCWN, 'SPW.. RlE.CE ,o eE RE.MOVEt> ...II C> ?..LAIi\< ~AN<,ES I NSTALLlt:, BE.FO'Rt Rtt\C.1'oR. S.Tli.R1 UP. 1

~-DG-'.0-H,2 ~,s4*RC.~;s-1so2 ~ ..r RC* \74*152. a_"-Rc.*\(n.7-152 4

(FM*I03e) VDSZJ*l5Y Pic:IMARY VENT PDT - PFtt.SSIJRIZE.R RE.1..\E~ TAJJK.. R:T. D. BYPM,,s DRI\IN,S l*RC-ik..-2.

-2~RC-\5"'.HS2 (F.J¥1*103S) ,~.---.--~

C.l.b~ l*CLIS~ t"vA-~-15'1 Dt:.TAILC 150 LB. S5 FL~E \ ~00 L.B SS FL~N<',E

\ SC.M.ED 40 SS Pl ?E.

r;,OQ 55BLI~ LB n rv-\j&H~'o,=ti. lJ J.:.ILMC: FL&. I r,o L!I ss !'L°" . I I 1-0A*P;4A~4B e 5PCOLPll:.C:t. SPOOl.PIE:.CL SPCO\.PIECIE... I RE.A,CTOR CO>JTA.\~ME:.!-JT DETAIL A. DETA.JL B DE.TA.IL(, 1 SUMP PUMP:!> 1 VENT AND DRAIN SYSTEM NOTt.*~ HOTE"~ NOTE',0:' L-------------- J UNIT I, SHEET 2 SURRY POWER STATION
FIG.9.7-3

A OCTJ?_. 191'()

OUTSIDE INSIDE CONTA.IN MENT CONTAINMEWT CFM-1008) FROM CONTA.1 NME.NT 150LATtON ';;YSTE.M SOV

~c~~~o~~-

l~'~VS-103-ISC (l=M-IOOB) VDS-15 ~ REACTOR COOLANi F='ll TEA.- *r fVG-106-152 FROM CONTAINME.NT ISOLA.TtON SYSTEM WITH PUMP INTERLOCK--....,. I I I V OG208B I -DG*22T* IS2

.. 2~-0G*'Z.4Z-152 CFM-IOOBl NOTES:

L..M,C.

~ ~-c.-.PPe.C CONt.J FOR MCliORlt,,IC!r.

OEIJOTES PE~E"TRA'TIO\..l J,J(), X~ DE.t..lOiE~ 'SUPF'LIEC Wl"'TH P\JH'IP~ A.LL PRES5oURE "'U*,P~(FT, Pl,PDt=P<:.V) TOBE ... A, SPOO\.l'l&'iS TO 8E. USEO FOP. DR-'ININ<'>-

.... HN" REKTOI\. IS SHUTOOWW. 51'00\.PIE.C.'E.

TO AUX'ILIARY BLDG. SUMP (11448-FM*IOOA) ffllE. R.El'<OVlol) I\Nl) IILI\Nl< 1'\.l'NCTI:: IIH,T/.LL'E.ll

!'19'011.£ 1!.U,C.TOII. 51'1\\\T IJP '~1"-DG-230-(.52

a*cvA-113-154 VGS*60B <°SA*II *IS+ CFM*IOOBJ FM-101,A 54*

-VDS-15l!f R.EFERENCE DRAWIJ/CS: *~

FLOW DIAGRAM COMPONENT COOLING WATE"~ SVS. Fl.l-22A&.8 FLOW OIAGfrAM SAMPLING .S'r'.S, FM- 32A

!! I FLOW DIAGRAM VENT&ORAINS f=M-1008 ~ FLOW DIAGRAM REACTOR COOLANT .SYS. F~-*193Ad,.B ..~' !:, FLOW Dl,A.GRAM R£S1tlUAL HEAT REMOVAL SVS. FM- I04A -O<l"'-'-"'="'-"'---'-"-!a---j!,l FLOW DIAGRAM CHEM. &, \IOLUl.46: CONT. 5YS. FM- IOSA,.B&<;- ~

FLOW DIAGRAM SAFETY INJECTION .SV'S. . FM - I 0&.A

R, FLOW DIAGRAM REAC. CAVITY PURIJ:'l(AT!ON 11448-.F'M,-11i\A cr FLOW DIAGRAM ~ENT ~ DRAINS SVS. 11448- FIi-- IOOB

'? ~

9 ii-1144S* f"fl.l- 1'22

\/A.LVC. 'S.TEM ~KOFf'S FROM CONTAINME:NT ISOLATION SY.STEM fM*IOJA FM*IO&A ~ WITH PUMP INTERLOCl _

VCS*15X 7 '

7 2"-DA*l24*IS2 FLOOR! TRCNCH DRAINS cf~ DA-I ZI- I 52 (F'M*IOOBl f-o - S-15l VDS-15Y

LA H OUTSIDE" INS1DE:

CONTAINMENT CONTAINMENT L~ (FM*IOOB) DG-206 VB'5*l5Y VALVE PIT LEAK 2-DA-P-1/1~ IB SAl="EG,UAR~i ARE:A VENTa DRAIN SYSTEM OFF' POT SUi\AP PUMPS 2-DC.-TK-2 UNIT2 SHEETi SURRY POWER STATION
FIG.9.7-4 OCT.15,1970 8001-V'-J.::I - 817<;'1 I fJ(,M-100,e)

~

1"-vG-101--rs2

~

INSIDE ours1oe CONTAINMEN'T COMTAINUE LO (FM*IOOA)

-vo:s-rsv LO ~.'~CH-4'24-*ISZ (Ft.il*\05C.) r-ru-201-1s2-2n- DG-21~-/62 -

2'iRC-S!i3*l502 r---t,>J---------""-4-...___ _ _2'"'*,--c,c;---21-,--,s-.,_.1._-,,---i----, (FM*\02,A,\

. CL1'504'e_L_1?2 /

2';0G*218*152 VCS-15X VDS-ISY ' R.C. PUMP STAND PIPES VALVE: 5":EM :.C/,.k.Of'F'!.5

*"' ~.~~ ~ ~

N N N

"'7 !!' ~ 7 N ...,n 0 "'cl "'...,

L__ _ -,jj ~* ~ .,, (FM- I03B)

e~ :J,(', :!l ~,:, "' "! "'

~ "' *\.?

0 ~ I 1* O 0 0' "'a 0 g I!!

~ ;t~c-:;-,so2 IH---........... ;IN ~~N ;;:N ~~

J:11 JTI Jo" L.:N

'Ja1

_,N -~"

J,<l J;tl

...1N ';',J

~I

_J PelMA.f::Y D12.Alt,J T£A.NSFER. TAJ.Ji:::

------ ---------(-F-M--1-0_4_A--)------- ~ )'i'-12C*:~-1so2 IH Ii,':. DG-247-152-- , VALVE STEM L'.2~0FF6 ;i~VG-112-,s~7 -:£1---~--~ ;;; eJ "'~ I "' .,.cJ: "' "'"'

, !'"'? "' ""'

0

'--;N .;

0

,J, 0

0 2"'

~

0

'i! "'~ .0 1P N ~

2-DG*P-IA8.1B PRIMARY DRAlN

~

LMC

"'J, '1 "! 7 "' ,?

T'RI\W.Sf'E.R PuMPS

-"~

l~5l*305-IS2~' 3000LB 0 D 1 0 Q 0 0 0 (FM-106B) Qi ,-,' i:: 5$ UN\Ot; (TYP)

'-,N l<'lkt '-ieJ :ni-r ~-~ '-\N ~

Ji' I Ji'I J;"I rt< Ji' Ji<l,-.J J11 rt<I ~I rt<I (FM-105C) 3000 LB SS UNIOh (TYP)

.. / (F'M*IODB) ==='""'---H 4*RC.-420~--------j ~

2."- le!L-ID3- 15.:2.

~ (IIAAEJ- FM- r 14 A) i-,,...-CG__-.__...

__ -,_sz...,.1-1 H ~ ~ f'.D<":,*l10-15Z f RC-445-\502

'-"--H~ z--RC-Ll59*152 (FM.-103B) \~is:,

2"-DG*ZJZ.-152.

\fc~ fE'. De!u EDTO MATCµ \ i;;oo Le, Fl.4 5

H I OtT A 1(-RC*40p*1602:: (FM-1091>.)

~ ISOLB \

LB FLC:. RJ2 ~ BUIJD W FL-6 r: R,C.* 40G,-l '5Q 2. I :::5FOOt... LO...-A.T/Ot,.J S0 -:sc:.H 4o ~~ '5C~ LO, e FM*103A-* A 10:. Ge e P(P8"------!_/

~ 1"2.n I

I P!PE--1_; A[OC:,. ~ B 12~ (TYP) I (TYP) 38 -- DETC NOT"E P....

'5POGL P(ECE DETB NOTE t...

DETA NOTE. A. VENTS DRAIN SYSTEM UNIT2 SHEET 2 SURRY POWER STATION
9.8-1 12-1-69

9.8 COMPRESSED AIR SYSTEM The Compressed Air System includes a .Service Air Subsystem, an Instrument Air' Subsystem, and a Containment Instrument Air Subsystem for each unit. Air service to certain common station areas may be provided from either unit.

9.8.1 DESIGN BASIS The Compressed Air System is shown in Figures ~.8-1 and 9.8-2. The design objective IOf the Compresse_d Air System is to assure ava_ilability of sufficient quantities of compressed _ai-r of suitable quality and at the pressures required for station operation *

Design pressures are dictated b'f t;he -expected uses of instrument or service air.

Design temperatures are those resulting from extreme ambient conditions and are based on 105° F service cooling water. The instrument air dew point has been 0 selected to be below the lowest indoor temperature expected (50 F) at the station location. The air receivers have been sized to provide breathing quality air and air to essential systems for 10 min after shutdown of all compressors following a loss-of-power accident. Design data for components of the Compressed Air System are given in Table 9.8-1. The instrument air compressors, instrument air receivers and drier, and piping,

valves and supports to critical instruments and controls are designed to conform with Class I seismic criteria (Section 15.2.1).

9.8-2 12-1-69 TABLE 9.8-1 COMPRESSED AIR SYSTEM DESIGN DATA SERVICE AIR SUBSYSTEM Service Air Compressor Number 2 (one for each unit) Discharge Pressure, psig 110 Discharge Temperature, °F 120 Capacity, scfm 373 Service Air Receivers Number 2 (one for each unit) Capacity, n3 280 Design Pressure, psig 125 Design Temperature, °F 400 Operating Pressure, psig 110 0 Operating Temperature, F 120 Material Carbon steel Design Code ASME VIII
9.8-3 12-1-69

INSTRUMENT AIR SUBSYSTEM

., /

TABLE 9.8-1 (Continued) Instrument Air Compressor Number 2 (one for each unit) Discharge Pressure, psig 110 Discharge Temperature, °F 120 Capacity, scfm 373 Receiver Volume, ft 3 , 36 Instrument Air Receivers Number 2 (one for each unit) Capacity, ft3 280 Design Pressure, psig 125 Design Temperature, °F 400 Operating Pressure, psig 110 Operating Temperature, °F 120 Material Carbon steel Design Code ASME'VIII Instrument Air Dehydrator Number 2 (one for each unit) Capacity, scfm 400 Dew Point at 100 psig, °F 35

Dew Point at 0 psig, OF Type

-12 Refrigerant

9.8-4 12-1-69 Instrument Air Aftercooler, TABLE 9.8-1 (Continued) Number 1 per compressor Type Shell tube Tube Side Flow, cfm (air) 273 Shell Side Flow, gpm (water) 6 Shell Side Design Pressure, psig 75 Tube Side Design Pressure, psig 135 Code ASME VIII Not Stamped Shell Material Steel Tube Material Admiralty Tube Sheet Material Service Air Aftercooler Chrome plated steel Number 1 per c9mpressor Type Shell tube Tube Side Flow, cfm (air) 273 Shell Side Flow, gpm (water) 6 Shell Side Design Pressure,-psig 75 Tube Side Design Pressure, psig 135 Code ASTM VIII Not Stamped Shell Material Steel Tube Material Admiralty Tube Sheet Material Chrome plated steel
9.8-5 12-1-69

TABLE 9.8-1 {Continued)

CONTAINMENT INSTRUMENT AIR SUBSYSTEM Containment Instrument Air Compressor Number 4 (two for .each wiit) Discharge Pressure, psig 100 Discharge Temperature, °F 325 Capacity, scfm 15 3 10.7 Receiver Volume, £t Containment Instrument Receivers Number* 4 (2 for each wiit) 3 11 Capacity, £t Design Pressure, psig 200. Design Temperature, °F 650 Operating Pressure, psi 100 Operating Temperature, °F 130 Material Carbon steel

*Design Code ASME VIII

\ 9~8-6 12-1-69 TABLE 9.8-1 (Continued)

Containment Instrument Air Drier Number 2 (one for each unit) Capacity, scfm 15 0 Dew Point at 100 psig, F 50 0 Dew Point at Atmosphere, F 50 Type Desiccant Containment Instrument Air Prefilter Number 2 (one for each unit) Capacity, scfm 40 Retention Size, microns Filter Element 10 Felt-wool ** Containment Instrument Air After Filter Number 2 (one for each unit) Capacity, scfm 15 Retention Size, microns 10 Filter Element Felt-wool COMPRESSED AIR SYSTEM PIPING AND VALVES - . Materials Carbon .steel, copper and bronze Design Code USAS B31.1
9.8-7 12-1-69

.9 .8. 2 DESCRIPTION The four (two per unit) 100 percent capacity air compressors of the Service and.

Instrument-Air Subsystems operating at 110 psig are located in the turbine building in a tornado missile and earthquake protected area. The instrument air compressors are used to provide air as required for instruments and contr:ols; associated with each unit, except for. instruments located within

the conta*inmen,t I

structures. However, the instrument air compressor also serves as a back1:1p to* the Containment Instrument Air Subsystem. The 110 psig air pressure in the headers will be reduced to that pressure required by the instruments and. con,trols at the point of use. The service air compressors are used to provide service, air at hose connections in each unit for operating equipment and tools during normal

operation and refueling. .

The service air receiver of each unit is cross-connected to the instrument air receiver of that unit to allow service air to be used as backup for instrument air. When low pressure occurs in the instrument air receiver, a pressure c~mtrol valve in the cross-connect piping between the two receivers allows air to flow from the service air receiver to the instrument air receiver. The service air receivers of each unit are cross-connected together and one service air compressor is normally in continuous service and the other on standby. A means of periodically alternating the service air compressors is provided.
9.8-8 12-1-69 The two service air receivers are cross-connected and the instrument air receiver' of each unit is cross-connected with its respective service air receiver; there-fore, each instrument air receiver is backed up by both service air receivers at all *times. The air compressors of the Service and Instrument Air Subsystems are of the nonlubricated type to eliminate oil contamination of the instrument air, and are all of the same size. Instrument air is dried and filtered, and is of breathing quality. The Containment Instrument Air Subsystem consists of two 100 percentj capacity nonlubricated type, air cooled, air compressors operating at 100 psig, with receivers and desiccant type air driers. This equipment is located in each containment structure and provides instrument air to all pneumatically operated instruments and controls located inside the containment. The receivers of the Containment Instrument Air Subsystem are initially charged to their full capacity and pressure (100 psig) prior to startup. During unit operation, the compressors draw the air needed to recharge their receivers from the containment atmosphere, thus balancing out the effect of air released by the pneumatically operated instruments, controls and valves. The Containment Instrument Air Subsystem is isolated from the Instrument Air Subsystem by check valves and containment isolation valves, w~ich are normally closed. The two compressors in each Containment Instrument Air Subsystem are sized to have one in continuous service and the other on standby. A means of periodically ~ alternating the compressor in service is provided.
9.8-9 12-1-69 .

The instrument ai.r compressors and the service air compressors are cooled by.

service water, (Section 9.9.1) which is available during normal unit operation

*and after a loss-of-power accident; the containment instrument air compressors are air cooled. The instrument air compressors and the containment ins.trumen.t I

air compressors and their driers are connected to the Emergency Power System (Section 8.5) so that continuous instrument air supply is assured after a loss-of-power accident. Station instrument air and service air lines penetrating the containment struct1,1res are provided with normally closed manual shutoff valves located out-side the containment to seal the containment internal atmosphere from the out-side atmosphere during an accident. Instrument and service air line penetrations I are isolated in accordance with Class V piping, as described in Section: 5.2.~ The equipment includes the conventional accessories, such as cylinder cooling systems, storage receivers, aftercoolers, and safety valves. 9.8.3 DESIGN EVALUATION

.The following devices are provided to preservE? an adequate instrument air supply under i!bnormal conditions and to ensure system reliability:

1. Service air backup to Instrument Air System. If instrument air pressure falls, a cross connection with a pressure regulator is provided between the instrument air receiver and the service air receiver to maintain

instrument air pressure. A by-pass line is also provided from the

9.8-10 12-1-69 outlet of the pressure regulator to the *outlet of the instrument air receiver. In the event 'of instrument air receiver failure, instrument air can be supplied directly from the service air receiver~

2. Service air backup between the two units. Two pressure regulators are provided between the service air receivers in each unit. If service air pressure falls in either unit, the other unit provides backup service air. This also results in each unit 1 *s instrument air receiver being backed up by the two service air receivers.

3. Instrument air backup between the two units. This is provided by means of cross-connecting lines between the two units at the.main headers.

4. Instrument air backup to Containment Instrument Air Subsystem.

event of the loss of both containment instrument air compressors and In the ). receivers, containment instrument air can be supplied from the instrument air system by opening the manually operated valves in the cross-connect line provided.

5. Check valves at the containment instrument air receiver outlets. In the event of the loss of one c9ntainment instrument air receiver, the valve closes preventing containment instrument air from escaping from the other receiver *

I

6. The instrument air compressors and the containment instrument air compressors and their driers are connected to the Emergency Power System

9.8-11 12-1-69

9.8.4 TEST AND INSPECTION Testing of this system is unnecessary because of its normal day-to-day operation.

Inspection is performed in accordance with normal station maintenance procedures

FIG. 9.8-1 E , OCT. 15,_1970

- \/6S*G.08 TURBINE' BLO(,i ---5ERVIC.& BL06 5E-k'v'KE- BLOC:! 1 :

--AUXILIARY BLDG / 'a;"

g

~

VGM-2-0~" ro UNIT,, CONTA.INME:UT

~~p,:iT~:~*=~~~ ~~Q;H~OA~:~~;~~

5EE: I'"* LJWG$ ,_.Of**, LOCATION~ Sl'Z~

'l.-*AC.C*8-'2.1B VGM-:?OS - ,~-ASC-3?*'21B I"* 5C-33-'2.I B !--------- l'~ASC- 30*21B- 0 1

t- z V65*60B

,*.YCM-20G fvoM-200-(TYP) "'0 ok P!C :,: z -VGl3-60.B 'c.-11.CC-107-'2.t.e r- 1"-N~,.-3e-211c SA-IOI Fl::OM Ut.JIT 2 ~ 1 VGM*20S- ..

VGM-205 . I1548 *F'M-15 A L.:~2_*-.A~SC_-_,o_s_-~2~,s-~-------C>I-<>/<>-' , \}"*AC:1>5)-21B l'-ASC-3S-218*

'1.'-ASC*<>*'I.IB r--~~~~----'--=-"---~~---~~~~-~~~~--,.~~1-----"~~-"---=--~~---~~~~~ 2'*A5(*5*21B I I ....~~~-1---'--'--,,--~-----ae>-"""-, \ \ .......C CL4'.:'*Y

_. / -- l'~ASC-31-219

-!*VDM*2DD(TYP) -~A'R.. *r:-:i lQ 'I"" RM NO.I \ ,[ E-L '2.1'-'=," -'2"A~C.*3*'21B .-f>..IR SUPPLY Ut:,1::: ~o~ 11,1-.. r-s EL G2'-o ~I IN LABS E-L 13 1*0" 1'N:C*4~-ZIB l"*ACC-'2.l-c;'.11:!, l~VOM-20D(TYP)*-:

VC.W*l5A EL SS'-Ei

/ . iO '>JtHJT ILATI ON A.IR S~JF-r-*e--,~l-lLkfOk.H,..(.l '-vc1-1~D ~- co IJj 'RO L.S. '-VGM-205 EL 11-o' ~

e,OIL ~- I*. I, -;, .I l"-A.C:. *::.. '-'2.lB "'" N~ f"-ASC-34-~l'B-'*

"'u' ...' '0 L...... VGM-20$

VGM-'20S-t=*ASC-54-1?:i~yi

'?..1*ACC*'24*?.IB TO HIGH LEVEL INTAKE 5TRUC..TUli:.E.

f'-ASC..*8-~\\ 1'-ASC-7*2\B - ~ VCM-10G- '2~-A~C*0<;)-'218

~-k,<1-l:,,+....,.t-'---~--'--'--------AIR. SUPPLY FOR OTH'ER

j. -VCM-1...0(J ,a*iAcC.-4'2.-(.IB INS1's IN SERVICE BLDG Uu

<.) %**LW*l'Z.l*IS2.

(SM-30A) f-voM-20D(TYP) VGM*'l.OS- - . - tll

~ a >-i::' FROM 2*~ACC-139-'21B AIR i"o EQUIP RM"'2.t~1 "'"' Pl l::,A-IO'Z. ~~-A~C.-G'2.*21 B, ~ < I- 2'i ~ (11548-FM-'2SA) fv"'M* ZOS R~ .

l"*AK31-~16 rvoM-WD(TYP) * ~ ~ ~~~ 5E:RVtl'.t AIQ SUP[.1LY St"R\ilCE* AIQ ~OR /P:M<;i-lOO-t4-iJ 1=A'S.C*3"*:21&

.VG.h1*'20'S IN5H?UME:NT Al"' ~ 1~ ...1_70J,i'*ZOD FOR AUX GE-N CU~Jc.LES HTG B'.'llLE-R ~J"'

A~TE-RCOOLC"'-' EL '3B'*G,* ,1 'z_ ~ VOM-200 j *VOM-200 -

.I' (TYP) l*IA..*f:*I IN5TQUME:NT AIR COMPRE:55GR 1 FIWtJ eE-AlclNCi-- ~4 *VOM*2.0D(TYP) 1-IA*C.,*1 (OOLING H,zO VG.M-ios 'Z."-A.SC-'3-'2.16 TO UNIT'Z.*

S&RVICE Al 12 HD~ (11548-Fl'l-2.'51'.)

...~~,A!Zx. Pf FROM UUIT 'l.. bEJ."VIC.E AIQ . 'Y-A!rC.* 93

P5 I'\_ *:_

f' J!t..'L,_l.-f.~<t-....ct,I Ali;.? TO

~E:C.EolVEi:l PQE5S COU\J ---;J~,,.!lo'.~*' .;....J......jj,i!!--.1 WAL l.. LOUV~E.S (II ~'IB -rlo\*:I.~ ... )

Fl>OM U~IT Z ~E~Vll& A(g, Q&!EIV'EP (11548 *FM-2!5Al

\- ....'5C. -ss- 219 -~~-----~--~,---(

'f I OPEN TO I IN NORTH YA.RD l*RMS*P- IS9 l~~.r~:

I

*1 AREA 1 \:..ASC-21-'G.16 COITT;i~~\i~~ ~~T'r,_FLATE , Y2,\iowl-WD a~Il~c:FJ. I \

I NO-TE5: ~ONTAMJNATIC?i 6LOG,

;*.f2~"' ~~ h *- IUDllATE-5. FURUIS.HE-0 BV 1:Q.UIPMl:UT MFR, PM!E:5:>URE- COUIJEC.TIOU5 ~*'uuLE55 IJOTED, ~f:>~2 TEMPC'RATUQE COUUElTIOU:> 11 UULl:!>SUOTE,O. ~...,. ..J IJUt.iBE-Q 51-1.0WU AT l<EAC.TOR C.O\JTAI lJME:IJT ~ ~ ~~ LIi.Ji: IIJOICATe:5 PIPILJG Pl:Wf::TRA.TIOU UUMBCR.

FROM BE/IRING COOLING H:0 svs.__..:,.__ _._____:....:::=:...:__ _:__ _ _ __J ~ z~ ij 3 AEFEl<E-UCE:- DW65: E-L G:.'2.1-01* t: I,) ;t. ::i ~ CO~T VAC.U'JM ~ LE:A.l<ACJC' NO\JITO~ 5V~ F-M

IO'ZA r-Asc-1CD-2.10-...-*

E-t.. -se.*-~d

~t ~1, if ~ !o*. ... BEARING COOLING; WATER SCHEDULE OF PIPING PENE.TRATl0iJ5 FM-2lA F\1- IA TO BEAR~NG ___:;::::;--- TYP AIR SUPPLY HOR COMPRESS[.0 AIR LINES ,11548-!='"M-2SA C.OOLING t-laO S'iS TO GROUP OF INST~ EL 131.a* EL 131.0* VENT *1 DR.~\1"'1 0Y$ F-M*IOOA IN TURBINE BLDG ---~*-voM,'2.0C (TYP) Whfol E' DIS~OSAL S"?. FM-'30A VGM-'?OS l"*A5C*18*2.18 2"To ~ I (F'M*27A) , - - - - - - - - - - - - ~

VGM*'Z.05 Al Q TO VE-IJTI LATIOW cmJT~OL5 l~*AC.(*44-'2.I B VOM-'200 VGM-2.0S

'2. 11 -ACC-'2.3-'2.IB VGM-'205 EL 13'-oJ-t*~--lXH l~Asc-n-2,e,..1 I ~--_..__2_*,_t.._c_c.,-3_<e>_-2_1_B_ _ _ ~ :I !

r-ACC-26-'2.l'B TURBINE BLOG-~ ! COM PRESSED Al R SYSTEM TO TURBINE- CONT. "lt-l.S l"R. J\\ I{ CONT, I~STR f>.IR RE.CE.IVE.R TRIP SY5TEM REACTOR CO>lTAINMEN'T ~ COW\PRE.$SOR (2 5T~GE.*. I* IA.-C-2A

-UNIT I - - -- - -- -- ,_ - - - -- SURRY POWER STATION

C K SF c FIG 9.8-2 OCT. 15, 1970 I AUXILIAAY SERVICE BLOG T:

.BLOG TOUIJlT.I 11448*FM

25A

~

z.":.ACC-120-IS'I C.L 151 I CL 2/B 2"-AC.C-12.S*'ll~ ~ - - - - - - T O '2"*ACC-39*21B (11448*fM-iSA) YG:;.,-Cooa RR RR

!tMS*fOO*C.O i

I VGS*GOB [. __ _ ~.--~.* Z"*ACC-107*21B I :..;;~,I ~ I Z.~-ACC-139*21B t'*ASC-105*'218 3." : OEHYD'."~A.,. J!-c 2-IA*D-I 3 VC1*15D 4'-A.CC.-102:-151 . V6S*&OB ~~ INST AIR RECEIVER ~~

;f ~

2-IA-TK-r

J..,... ______

I I I I I I

'2."*A':.C*1'2b-ZIB YUM*105 I +-----~=~---~==~----------~

ro 2*-As(-'2G*'ZI 8 l>J~TRUMENT AIR i

r "l"'" COMPRESSOR FROM BE1'RIIJG 2

IA-C- I caoL11..t<l. \-l.zO s.v~ ,

TO UNli Al R TO V£:LJTJ LAT IOU~' Z"-'A':,C-10.5-ZIB. C.Ol;JTROL5

SERV1ce AIR

_HEADER -l'~'A.CC-'4S*'21B (1144B-FM-'tSA) AIRTO WALL LOUVE-RS

&A-200 NOTE' LMC - LEAKAGE MONITORl~G CONN lf INDICATES FURNISHE:b 8Y EQUIPMENT' MFR 1

FROM UNIT I PRESSURE CONNECTIONS 3/4 UNLESS NJTED SERVICE AIA VC!-150 TEMPERATURE CONNECTIONS!" UNLESS NOTl::D R.EtEI\/ER NUMBER SHOWN AT REACTOR CONTAlliHE°tfT LINE_. 4'-ASC* 102- \SI

/1144!>-F ... z..,..) INDICATES PJPING PENETRAT,ION NUMSER*

REFERENCE DRAWINGS:

BEARING COOLING WATER

-$.C~[OULE Of' PIPING PENETRAT'!°oNs*

FM-~3A F>1.*l4 COMPRESSED AIR LINES l!r44B*FM-aSA CONT VA.C ( LEAKAGE MON \TO~I NC. 5'1~ FM -I01A

~~-'-VOM-ZOD

CONT. INSTR. AIR DRYER PREFIC\ER '2-IA-~-2.

. 2" -ACC.* I0~-21 B "'

2"

' 0 TO FER':'£>>.LNEL AIR \.. k i~ l\11lER ,:,OCR LOCKING >3> RIN t>EV\CE r~ACC-149-2\B 7 5£:R'VICL AHl fRO~ B'E.AR!MG. COMPl=tESSOR REACTOR HEAD ..........._-{'. .s,oRAGE AREA COOLUJG. HzO SY~ '2*SA*C* I ,--~---~---~~--------, VOM*20D 2"-ACC- JOO *21B , .. ~-ACC-\04- lB 10 '\/ENT TO BEARIIJG.

TYPAIR SUPPL.'r HOR FAN CAMP l< [FB-<>Al coOLIWti HzD'~ys TO C::.ROUP 01=- llJ~T'S I~ TURB BLDG i.:. VGM-Z.OS VGM-205 t-..--+-,:,<1 tJ~-ii?o~~TI LATIOLJ

2"-ACC-12'3*'21B VOM-200 fAsC-ll7-21B, VGM*'205--*

VGM-2.05 rn;~:01 (FM-27A) C.0\...11. J.1-l'i!,IR. Al R COi-Ji. INST?. Al;a

i-\lOM-20D 1~-Acc-12,-21e.---- 1 . COM PRE'5~0R (TYP) C:OMPRESr.R

<z-sT.a.GE.

t-lA'C- A (-i;. ETAG~J

~-IA*C-'l:B COMPRESSED AIR SYSTEM

.~* TO TURBINE CONTROL TD TURBINE UNIT 2, SHEET I VALVE TEST CONN S TRIP SYSTEM TURBI~ E BLOG REACTOR CONTAINME~T d SURRY POWER STATION
9.9-1 12-1-69

9.9 SERVICE WATER SYSTEM River water is the source of service water for the Surry Power Station. Since this water is brackish, it is not directly used for cooling critical equipment.

Service water is used as cooling water for heat exchangers which remove heat from the Component Cooling Water System (Section 9.4), the Bearing Cooling Water System (Section 10.3.9), the Recirculation Spray System (Section 6.3.1), and other station applications such as air conditioning, chilled water, and makeup to the flash evaporator. The Service Water System is shown in Figure 9.9-1.

9.9.1 DESIGN BASIS The Service Water System is designed for the removal of heat resulting from the simultaneous operation of various systems and components of two units based on a maximum river water temperature of 95° F. This temperature is two degrees warmer than river model tests indicate for the river water temperature on record. The Service Water System is designed as a Class I system (Section 15.2.1)

__J

*9 .9 *. 2-1 12-1-69

9.9.2 .DESCRIPTION Service :water is supplied from the Circulating Water System (Section 10.3.4) by gravity flow ':between the high level intake canal and .discharge canal seal pit. During normal operation, the water level in the intake canal is app,roxi-mately 20 ft above the level in the seal pit at the discharge canal. This differential head supplies the service water to par.allel flow. loops through the bearing coolin,g water heat exchangers, component cooling heat exchangers and recirculation spray heat exchangers which are also .in parallel with the main co.ndenser ..

I Remotely operated ,butterfly ;valves are installed at the four inlets and out-lets of' each main condenser and iri *the supply lines to *the bearing cooling water heat exchangers, the component *cooling heat exchangers, and the recir-culation spray heat exchangers. The operation of these valves is listed in Table 9.9.,,-2. These motor .operated valves are positioned automatically for various incident conditions, to conserve approximately 25,000,000 gal of water in the intake *canal for critical services. Power fo.r these valves is nonnally from the station. power supply; however, in the eveht of a loss of power, provisions have been made to s:upply the valves

. *with power from the emergency generators.

The .maximum service water requirements of*the system during normal operation are tabulated as follows:
9.9.2-2 12-1-69 Component Cooling System Flow, gpm 18,000 Heat Transfer, 106 Btu/hr 100. 6

No. of Exchangers Operating 2 (one for each unit)

Furnished 4 (2 for each 1,1nit) Bearing Cooling System 24,000 54 2 (each 3 (each unit) unit)

Chilled Water Condenser 2,185 15. 3 2 (one for 2 (one for each unit) each unit)

Control Room Air Condi-tioning 300 1.4 1 3 Charging Pump Lube Oil Coolers* 90 6 (3 for 6 (3 for each unit) each unit)

*Flow to the chilled water condenser is Supplied by a

pump which takes its suction from the circulating water system.

9.9.2.1 Inspection and Testing The Service Water System is in continual use and requires no special testing. 9.9.2.2 Accident Design Basis During a loss-of-coolant accident without a loss of station power, the isolation valves to the recirculation spray heat exchangers open in the affected unit. All valves in the service water supply to the other heat exch~ngers will remain open. During this type of accident'the service water requirements ~11 increase above those listed under Section 9.9.2 an~ includes the flow to the recirculation heat exchangers which is as follows:
9.9.2~3 12-1-69

Recirculation Spray System Flow, gpm*

24,400 Heat Transfer,

106 Btu/hr 222 No. of Exchangers Each Unit Operating 4

Furnished' 4 If a total loss of station power occurs simultaneously with a loss-of-coolant accident in either unit, the recirculation spray heat exchangers isolation valves open and all other isolation valves in the Service Water System are closed. Under thes.e conditions the maximum service water flow 'to the recircu-latio~ spray heat exchangers will be 12,000 gpm. In the event of a total. loss of station power only, the recirculation spray heat exchangers isolation valves remain closed and all other service water isolation valves remain open

The operation of the valves is listed in Table 9.9-2

9.9.2-4 12-1-69 TABLE 9.9-2 AUTOMATIC OPERATION OF CONDENSER AND SERVICE WATER VALVES Initial Valve Action Service Main Condenser Accident Water Value Valves Desig~ Basis Accident (a) Open recircula- (a) Close all Loss-of-coolant either unit tion spray heat valves, both and loss-of-station power exchangers to the units affected unit (b) Close all others Loss-of-coolant either unit (a) Open recircula- {a) All remain with or witho~t a loss of tion spray heat open, both power to affected unit exchangers to the units affected unit (b) All others remain open Total loss-of-station power (a)

(b) Leave recirculation spray heat exchangers closed, both units All others remain open {a) Throttle all valves, both units
9.9 . ~!2-:5 12- 1--.6:9 1

9.9.2.3 *Emergency Serv:ice Water Pumps In the ev:e.n:t of a loss *of station ,power at the river intake, three .die.set driven, vertical emergency .se~ice *water pumps have been provided for *bo,th units at the river intake structure to supply makeup *to the high level .cana'l...

One. of the ,pumps is also .supplied with a dual dri:ve *arrangement ,:Which cons:ia;t.,s of an elec*tri.c motor in ,addition t.o *the diesel driver, and will be used dur.i~g

*normal s:t.ation ,maintenance. Each pump is *sized for .a *flow of 15 ,0.00 gpm .s:and :a total head ;pf 4'5 .'f.t, ,and ea.ch .can s.upply sufficient *w.at.er to contr.ol any ,.of the :acci.d.e.nts *us ted .in Table 9 .:9"".2.

The Jollowing cri;te.ria .were us.ed *in i;izing .the emergency serv*ice water pumps . ::

1. Des~gn JJa.sis Ac.cident (los~-,of-coo.lant ace.ide.nt ,.in .either unit -.and a

.,tot.al .l~s.s-,of-_s:ta,tion :power). *the maximum .wa:ter :flow *required is 12,000 ,gpm ,to the .recirculation spray heat exchangers. This .condition_.,

assuming \t*hat *the ,unit with -.only a loss-of-s*t*ad,on power .*remains on hot ,-stan.-d~Y-, would require .one emergency ser:v:ice water pump in operatio.n.*

-One PUJJ\P ;is .ass.urned to 'he .undergoing maintenan.c.e, .and one pump is assume.cl :to ,undei;:go _an ac*tiv_e failure .* .2. Loss-of,-coo"lant *accident .and a total loss-of-.sta*tion power, with the reqµirement that .the unit that d.id not unde.rgo the loss-of-coolant accident :must also be .cooled down. The-maximum flow wou1d be 21,000 ,gpm and would require *two of the .three pumps to be .operated

9.9.2-6 12-1-69

3. Loss-of-station power in two units. The total flow of 18,000 gpm would be required to cool down the units. This would require two of the three pumps to be operated.

9.9.J;..1 12-1-69

9.9.3 DESIGN EVALUATION In the event of the Design Basis Accident, the valves in the supply lines to the bearing cooling water heat exchangers and component cooling heat exchangers may be re-opened remote-manually from the Main Control Room if. service water to these systems is considere*d necessary.

The piping and equipment movements at the recirculation spray heat exchangers have been analyzed in accordance with the earthquake design criteria and have been installed to assure that no undue forces are exerted on piping or equip-ment nozzles.

The gravity flow of service water from the intake canal assures adequate cool-ing water to* the recirculation spray heat exchangers for more than 24 hr in the case of the Design Basis Accident, if water is not used for component or bearing cooling. This supply of cooling water is based on one unit operation with two half-capacity recirculation spray heat exchangers. It is expected that before *the end of this period, normal power will be restored to the circulating water pumps to replenish the intake canal water supply. Three diesel driven emergency service water pumps are provided in the event that power is* not restored within the .24 hr period.

If the second reactor unit is being refueled and the reactor head is not in place, continued core cooling is required, in the event of a loss-of-station power. Service water flow to one component cooling heat exchanger will be
9.9.3-2 12-1-69 continued and the cooling water supply in the intake will last for approxi-mately 12 hr without using diesel driven emergency service water pumps. A 4,800 gal diesel fuel oil storage tank provides sufficient fuel to operate all three pumps for 125 hr. One diesel driven pump is also electric motor driven. All pumps can be started from the Main Control Room through battery power supervisory channel. The electric motor driven pump is normally used to maintain the intake canal full of water when both units are shut down. The possibility of leakage fr.om the reactor containment into the service water through the recirculation spray heat exchangers after a ioss-of-coolant accident is discussed in Section 6.3.2. 9.9.3.1 System Reliability

A double set of normally closed paral.lel motor operated valves control the service water supply to the recirculation spray heat exchangers, thus pro-viding positive assurance that cooling water can reach the exchangers in the event of a malfunctioning valve.

Three diesel driven emergency service water pumps are furnished to accommodate the remote possibility that normal power cannot be restored within a 24,hr period. These pwnps will furnish water to the intake canal during the period when station power is not available.
9.9.3-3 12-1-69

Each pump is capable of supplying sufficient water to control any of the accidents listed in Table 9.9-2. Batteries required for starting the diesels are maintained fully charged by a trickle charge system.

9.9.3.2 Malfunction Analysis Failure of the Service Water System is precluded as follows:

1. Malfunction of the butterfly valves in supply lines to recircu-lation spray heat exchangers. A double set of valves in parallel are provided to ensure that water will be available at all times.

2. Failure to restore power to circulating water pumps after 24 hr period of stored water in intake canal. Three diesel engine driven emergency service water pumps are provided - each pump having sufficient capacity to supply water to control any of the accidents listed in Table 9.9-2.

9.9.4-1 5-1-71

9.9.4 TESTS AND INSPECTIONS The service water system valves in the lines supplying the Recirculation Spray System heat exchangers .are tested prior to initial unit startup to verify proper interlocking for various simulated accident conditions.

Periodic testing will confirm that proper operation of the valves is maintained. The design head capacity characteristics of the service water system were verified by determining flows through the Recirculation Spray heat exchangers during startup testing. The diesel driven emergency service water pumps will be tested monthly to

assure availability when needed. In addition, one diesel driven pump will be operated during tornado warning periods or at any such time it is felt that the backup operation of this pump materially contributes to the safety of the station.

The starting batteries for the diesels will be periodically checked for specific gravity and individual cell voltages. An equalizing, or overvoltage, charge will be applied long enough to bring all cells up to an equal voltage. Over a period of time, these tests will reveal weak or weakening trend in any cell, and replacement will be made, if necessary

B C SF FIG 9.9-1 V12-V\J..:l-9vv 11 FM-1'?.~A OCT. 15, 1970 l---

IH'. *-*!,:"I s':.wBTD-2-lSI R ~ .'JWI00-20 ./:,wizO e:LOWt!OWN TA!--lr', I::'i=l,AIH

- - - - ________ __LSW120 **--- -*--* _____ _ '-1Hl5 EG:itllF"MENT TO SE. \.OC>.TEO i ,i,,T THE SE.AL Pl"f DISCHARGE TUNNEL t u*t-.i1T I REc1R-C."i.Jl.j.;;T10'N 5PR~'i' 'NITI-HN GONT1',.1N~rni - - - "

HE,;..T--EiC.H1>,.1,HfE'iS. --,.ri j 1-RS-W, >RS*E*IS 1-R$-E*IC,

/ 'v) l-~5'-E-10

SV ,.

I ( '~Ii) _(_, ~"\ , " 1

+-.=--4VOT*20C:, , (T'iP) Cl.IOI~

H r ,-, ROM AIR COtlD UutT(?8*41A)

°. r,,." (FP-71A) ~1 t*~ ' ' I I r**i !_-4 µ i' Q

TO 5LUl([ TREl'sC. H

~_I MOV ~

3 l - L ~ )~&T *20E 1 G~ 3~ "~D-IZ.\ j ., f'ROh\ DE.MINEAA\.IZ:f.O

. (t'f?) \lll..TE.:S AR~A '=UMP PUMP -VVF-SA ";;"-WCW- I* l8b (FM*2'7A)

(TYP) 0 A-VfS-C..1*13" MOV

)WIOSA.

REJ*l\(TY~/

./ ~

I FROM VAC PUMP l-lEAT E'1.CU(i=M*(4A) fro RAD. t MONll"OR (fM*IZbA l (WP) Pl VOT*ZOG (l'll'J 5WI04A. 1-C.C-E.-IA. MOV Rl:J-IA-; . ~ _// ,:~~B I . a . GIRCULATJNt:. WATER PUMPS FE CAP/\ClTY

210 1000 GPM SW\OSB Pl HEAJ>*ZS'TDH 1><1,.!=,r-+----,~'4..1:SW\OOC. (fSIZEJ f(TYP) NOTE COND CCNN NO~

RTD 47,4B,51.f! 52. TO BE SWIOSA. TT If 150 LB FLG.

'i:.WIOS~ Tl FOR CONOENS~ R. l"U'o'e. NO 46 TO BE J" 150 LB 5Wl0SA CLEANING 'SYSTEM (T"YP). FLG.

FM*l\2A 0 UNIT l t.2

'24~MH COtv1PONENT COOL! NG H1.D VVf:"-51\ 1-CW*P*\D HeAI* E'JCCHA NGERS -2.4°MH 42"-ws-13,-10 42"-ws-12.10 UNll I L ____ _

1 BEARING COOLlNG H'l.o VVF*SA-"f , HEAi EXC.HAt.JGERS ~ (TYP) /

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9.10.1-'-1 12-1-69

9.10 FIRE PROTECTION SYSTEM The Fire Protection System is designed to furnish water and other extinguishing agents for fire protection throughout the station.

9.10.1 DESIGN BASES The Fire Protection System is designed in accordance with the standards of :the National Fire Protection Association and is based on the recommendations of the Nuclear Energy Property Insurance Association to provide the following:

a. Supply of water for fire fighting

b. System for delivery of water to potential fire locations

c. Automatic fire or smoke detection in the more critical areas

d. Fire extinguishment by fixed equipment activated automatically or manually

e. Manually operated portable fire extinguishing equipment at strategic locations

f. Fire barriers The following elements are designed to Class t criteria (15.2.1):

a. Engine driven fire pump

b. Diesel oil tank for .engine driven fire pump

c. Yard hydrant piping system Fire rrotection System design data are given in Table 9.10-1.

9.10.1-2 12-1-69 TABLE 9.10-1 FIRE PROTECTION SYSTEM COMPONENT DESIGN DATA Fire Pumps Number 2 (1 motor and 1 engine-driven) Type Horizontal Centrifugal Motor Horsepower, hp 185 Engine Horsepower, hp 185 Capacity each, gpm 2,500 Head at Rated Capacity, ft 244.9 Design Pressure, psig 175 Design Temperature, °F 80 Seal Material Pump Casing Packing Cast Iron Shaft Steel Impeller Bronze Earthquake Design Class I (engine-driven pump only)
9 .10 .1-3 12-1-69*

Pressure Maintenance Pump TABLE 9.10-1 (Continued)

Number 1 Type Horizontal Centrifugal Motor Horsepower, hp 4.5 Capacity, gpm 30 Head at Rated Capacity, ft 252 Design Pressure, psig 125 0 Design Temperature, F 90 Seal Packing Material Pump Casing Cast Iron

Shaft Impeller Carbon Steel Bronze Hydropneumatic* Tank Number 1 Type Cylindrical, Vertical Capacity, gal 4 7,5 Design Pressure, psig 200 Design Temperature, °F 100 Material Carbon Steel Design Code ASME VIII

9.10.1-4 12-1-69 Fire Pump Oil Tank Number TABLE 9.10-1 (Continued) 1 Type Flat Oval, Horizontal Capacity, gal 300 Design Pressure Atmospheric Design Temperature, °F 90 Material Steel Design Code ASME VIII Earthquake Design Class 1 Water Storage Tank Number Type Capacity,* gal 2 Cylindrical, Vertical 250,000 reserved Design Pressure Atmospheric Design Temperature, °F 5 Material Carbon Steel Design Code NFPA No. 22
9.10.1-5 12-1-69

Air Compressor TABLE 9.10-1 (Continued)

Number 1 Capacity, s cfm 8.11 Discharge Pressure, psig 100 0 Discharge Temperature, F 488 Low Pressure Carbon Dioxide Storage Tank Number 1 Type Cylindrical, Horizontal Capacity, tons 15

Design Pressure, psig Design Temperature, °F Material 363 0

- Steel Design Code ASME VIII

9.10.2-1 12-1-69

9.

10.2 DESCRIPTION

That portion of the Fire Protection System that uses water consists of water storage tanks, pumps, piping, hydrants, hose stations, deluge, and sprinkler systems. The two water storage tanks each have 250,000 gal reserved for the Fire Protection System. Well pumps furnish water makeup to these tanks which also serve the Domestic Water System. Two horizontal centrifugal pumps located in a heated fire pump house provide flow to the fire fighting equipment. One pump is motor driven and equipped with automatic control starting at 90 psi on decreasing line pressure. The second is diesel driven and equipped with automatic starting at 80 psi on decreasing line pressure. Each pump delivers 2,500 gpm at 100 psi discharge pressure

System pressure is maintained continuously by a pressure maintenance system consisting of a jockey pump, a hydropneumatic tank with air compressor and related controls and accessories. Automatic fire detection devices are actuated by exceeding a fixed temperature, a rate of rise of temperature or both, or ionization. Hose rack stations and portable fire fighting equipment*

are provided at strategic locations thrpughout the facility. The station fire fighting equipment is as follows:

a. A 12 in. yard loop with 11 yard hydrants strategically placed around this loop

_J
9.10.2-2 12-1-69

b. A manual or automatic water spray deluge system is provided for the following:

1. Hydrogen seal oil unit

2. Oil purifier unit

3. Turbine oil reservoir and coolers and high pressure fluid reservoir.

4. Main power transformers

5. Station service transformers

c. Automatic type water sprinkler systems are provided for the following locations:

1. 2. Turbine room floor areas including turbine oil storage room Warehouse

3. Auxiliary boiler room

4. Office records room

d. The CO2 protection systems consists of a refrigerated low pressure storage tank, distributor piping and controls for automatic or manual

'

I

discharge of COz. Smoke detection is provided in normally unoccupied areas where a high concentration of .electrical cabling exists. Auto-matic actuation of the CO2 system be means of smoke detectors with monitored lockout for tunnels only is provided for these areas. In

9.10.2-3 12-1-69 other areas protected by co 2 the high temperature detectors signal the annunciator and manual remote push buttons actuate the flow of co 2

High pressure systems are provided at remote locations.

Total flooding carbon dioxide is provided for the following locations: Automatic

1. Cable vault rooms and tunnels

2. Switchgear rooms

3. Cable tray areas Manual

1. Charcoal ventilation filter assemblies

2. Emergency diesel generator rooms

3. Turbine - generator enclosures High Pressure

1. Fuel oil pump room

2. Intake structure oil storage room

e. Manually operated portable fire extinguishing equipment is provided throughout the station.

f. Fire barriers are provided in the turbine building to cut off Unit 1 from Unit 2 below the turbine operating floor and between all main and reserve

station service transformers

9.10.3-1 12-1-69

9.10.3 DESIGN EVALUATION A reliable source of water is provided by two 250,000 gal water storage tanks which are maintained at full capacity. An electric motor driven fire pump backed up by a diesel driven fire pump ensures the delivery of water. Automatic sensing devices detect smoke and fires in areas where they might occur. Auto-matic or manual Fire Protection Systems provide a means for extinguishing fires should they occur

9.10.4-1 12-1-69

9.10.4 TESTING AND INSPECTION The fire pumps will be tested prior to station operation and periodically thereafter to check their capacity. Further inspections and tests will be performed to ensure the availability of the system .

SF FIG. 9.10-1 OCT. 1969 I *

C D E H K

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9.11.1-1 12-1-69

9.11 WATER SUPPLY AND TREATMENT SYSTEMS 9 .11.1 WELL WATER SUPPLY SYSTEM The Well Water Supply System provides makeup water to the fire protection and domestic water storage tanks, the hydropneumatic tank in the Potable.

Water System, the Flash Evaporator System and the Fire Protection System. The Well Water Supply System is shown on Figure 9.11-1. There are two cased water wells located south of the site having depths of 418 and 420 ft about 1,000 ft apart, and indicated as wells "B" and "C" as shown on Figure 15.1-1. Each well has a 200 gpm submersible pump

discharging to a well water storage tank. Each well pump has a separate underground discharge line which is interconnected at the storage tank.

Centrifugal type well water transfer pumps deliver water from the storage tank to consuming systems as required. The Well Water Supply System is designed to be automatically or manually controlled .
9 .11.2-1 12-1-69 9 .11.2 DOMESTIC WATER SUPPLY SYSTEM A 4,000 gal hydropneumatic tank, located in the fire pump house, is provided

, for the Domestic Water Supply System. Pressure in the hydropneumatic tank.

is *maintained at 40 to 60 psig by a pressure system, consisting of a pressure-level regulator, air compressor, and related controls and a

.accessories. Hypochlorinator equipment provides a means of chlorinating /

the domestic water supply. Piping from the hydropneumatic tank supplies cold water to safety showers, drinking water coolers, hot water storage tanks, and domestic cold water throug~out the station

9.11.2-2 12-1-69 TABLE 9. 11. 2-1 DOMESTIC WATER SUPPLY COMPONENT DESIGN DATA Hydropneumatic Tank Number 1 Type Cylindrical, Horizontal Capacity, gal 4,000 Design Pressure, psig 100 Design Temperature, °F 100 Material Carbon.Steel Design Code ASME VIII Water Booster Pump Number Type Motor Horsepower, hp 2

Centrifugal, Inline 15 Capacity, gpm 300 Head at Rated Capacity, ft 139 Design Pressure, psig 135

. 0 Design Temperature, F 90 Seal Packing Material Puni.p Casing Cast Iron Shaft 316 Stainless Steel Impeller Bronze

9.11.2-3 12-1-69 TABLE 9.11.2-1 (Continued) Air Compressor Numper 1 Capacity, scfm 8.11 Discharge Pressure, psig 60 0 Discharge Temperature, F 380
9.11.3-1 12-1-69

9 .11. 3 FLASH EVAPORATOR SYSTEM The Flash Evaporator System is shown on Figure 9.11-2. The system consists of equipment for the production of high purity water by deaerating, evaporating, cooling, prefiltering, demineralizing, and postfiltering well water or service water for makeup to the various station systems.

Supplem.ent~ry chemical feed equipment is provided for feedwater conditioning chemicals. High purity water is pumped to the primary water storage tanks (Section 9.2) for-reactor plant makeup and to the condensate storage tank for secondary plant makeup. The flash evaporator and the polishing demineralizer are designed to operate automatically after manual initiation. Evaporation is started manually on low level alarm from the condensate

storage tank and proceeds until an alarm is actuated indicating high level in the condensate storage tank. On high effluent conductivity another alarm is actuated and the ev~poratot effluent is blown down to waste.

The flash evaporator is designed to provide a normal net distillate output of 125 gpm when using 6th point extraction steam and a maximum net output of 220 gpm when using steam supplied from the Auxiliary Steam System. The prefilter is of the cartridge type. The cartridge type postfilter unit is designed to remove any suspended material larger than 5 microns in the demineralizer product stream

9.11.3-2 12-1-69 The flash evaporator, with an integral vacuum deaerator, is constructed for a test pressure of 22.5 psig to evaporate well water and service water and deliver distillate with less than 7 ppb of dissolved oxygen and less than 25 ppb of dissolved solids (as Na) excluding iron and copper.

The demineralizer vessel is a rubber lined steel pressure tank with Type 316 stainless steel internal distributing underdrain and constructed in accordance with the ASME Code for unfired pressure vessels with 100 psig working pressure. The vessel was tested hydraulically at 150 psig and loaded with a mixed cationanion exchange resin. The unit is designed to produce from a minimum capacity of 50 gpm up to a maximum of 460 gpm of reactor grade demineralized water. The Fire Protection System is discussed in Section 9.10.
9-.ifL 3-3 -. TABLE 9 .11. 3-1 FLASH EVAPORATOR SYSTEM COMPONENT DESIGN DATA 12-1-69 Evaporator Condenser Number 1 Type Shell and Tube

-Capacity, lb per hr 61,000 Design Pressure, psig Shell -14.7 to 15 Tube 750 Design Temperature, 0 F Shell 300 Tube 300

Material Shell Tube Carbon Steel 90-10 CuNi Design Code ASME VIII (Tube side only)

9 .11. 3-4 12-1-69 TABLE 9.11.3-1 (Continued) Auxiliary Evaporator Condenser Number 1 Type Shell and Tube Capacity, lb per hr 87,500 Design Pressure, psig Shell -14.7 to 15 Tube 125 Des i gn Temperature, OF Shell 300 Tube 300 Materials Shell Tube Design Code Carbon Steel 90-10 CuNi ASME VIII (Tube side only)
9.11 .. 3-5 12-1-69 TABLE 9.11.3-1 (Continued) Recirculating Steam Heater Number 1 Type Shell and Tube Capacity, lb per hr 2,880,000 Design Pressure, psig Shell -14.7 to 15 Tube so Des i gn Temperature, OF Shell 4po Tube 400 Material Shell Steel Tubes 90-10 Cu Ni Design Code ASME VIII (Tube side only)
9 .11. 3-6 10-15-70 TABLE 9.11.3-1 (Continued) Makeup Heater Number 1 Type Shell and Tube Capacity, lb per hr 145,800 Design Pressure, psig Shell 125 Tube 125 Design Temperature, of Shell 200 Tube 200 Design Code ASME VIII Recirculating Pump Number 2 Type Centrifugal Motor Horsepower, hp 100 Capacity, gpm 5,800 Head at Rated Capacity, ft 55 Design Pressure, psig 150 Seal, Type Mechanical Material Pump Casing Cast Iron Shaft H.T.C. Steel Impeller Bronze
9 .11.3-7 10-15-70 TABLE 9.11.3-1 (Continued) Distillate Pump Number 2 Type Centrifugal Motor Horsepower, hp 50 Capacity, gpm 465 Head at Rated Capacity, ft 240 Design Pressure, psig 300 Seal, Type Mechanical Material

Pump Casing Shaft Impeller 316 Stainless Steel 316 Stainless Steel 316 Stainless Steel

FIG. 9 .11-1 DEC.1,1969

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FIG.9.11-2 OCT. 15, 1970 V'L2-ll'J.:l-817171 I EVAPORATOR -""*WEF-12 *151 vrnr CONDENSER l*Wi-E.-'Z. 1l/'vtt,,1\TO t..TMO~ 1l(TYP) J , (TYP) TW

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9.12-1 12-1-69

9.12 FUEL HANDLING SYSTEM The Fuel Handling System provides a safe, effective means of transporting and handling fuel from the time it reaches the station in an unirradiated condition until it leaves the station after post-irradiation cooling.

The system is designed to minimize the possibility of mishandling that could cause fuel damage and potential fission product release. The Fuel Handling System consists basically of:

1. The reactor cavities, one in each unit's containment structure,

which is flooded only during unit shutdown for refueling, and a manipulator crane for each unit.

2. The spent fuel storage pit, which is maintained full of borated water and always is accessible to operating personnel, and a movable platform with hoists. It is shared by both units.

3. The Fuel Transfer System for each unit which consists of an underwater conveyor that carries the fuel from the reactor cavity, through the containment wall and into the spent fuel storage pit

9.12.1-1 12-1-69

9.12.1 DESIGN BASES The Fuel Handling System and areas comply with the appropriate criteria as discussed in Section 1.4. The applicable criteria are:

Criterion 4 Sharing of Systems Criterion 18 Monitoring Fuel and Waste Storage Criterion 66 Prevention of Fuel Storage Criticality ------ Criterion 67 Fuel and Waste Storage Decay Heat Criterion 68 Fuel and Waste Storage Radiation Shielding
FIGURE 9.12-1 2-lJ j---~- 1

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SEC\ION "f>..-6 FROM 'tf Dv'l/6 IIOE.Z.." FUEL TRANSFER SYSTEM
9.12.2-1 5-31-71 9.12.2 SYSTEM DESIGN AND OPERATION 9 .1:!. 2, l System Description Each reactor is refueled with equipment designed to handle the spent fuel under water from the time it leaves a reactor until it is placed in a cask for shipment from the site. Boric acid.is present as required in the water to ensure subcritical conditions during all phases of the refueling process. In each reactor cavity, in each unit's containment, fuel is removed from the reactor vessel, transferred through the water and then placed in the Fuel Transfer System by a manipulator crane. There is a separate Fuel Transfer System with each unit. It is then transferred through the fuel transfer tube to the spent fuel pit, Fuel is removed from the Fuel Transfer:_. System and placed in storage racks with a long manual tool suspended from an overhead electric monorail hoist on a bridge structure mounted on a movable platform that runs over the new and spent fuel storage.areas, which are connnon to the two units. After a sufficient decay period the fuel is removed from storage and loaded into a shipping cask for removal from the site. New fuel assemblies are received and stored in racks in the new fuel storage area. The new fuel storage area does ~at contain any water. N~~ fuel is delivered to the reactor by transferrin!l it from the new fuel storage area to the spent fuel storage pool and taking it through the transfer system. The new fuel storage area is sized_ for storing two-thirds of a core as detailed in Table 9.12.2-1, A portion of the fuel for the initial

core loading is temporarily stored in the spent fuel pit

9.12.2-2 12-1-69 The reactor cavity and spent fuel pit are reinforced concrete structures with butt welded stainless steel plate liners. These concrete structures are designed as Class I to withstand the anticipated earthquake loaqings.

A+l liner butt welds conform strictly to the requirements of Section IX of ihe ASME Boiler and Pressure Vessel Code and are provided with test chambers to check for leak tightness. Fue+ handling data are given in Table 9.12.2~1.
9.12.2-3 4..,15 ... 70 TABLE 9.12.2-1 FUEL HANDLING DATA New Fuel Storage Area (common to both units) Core storage capacity 2/3 + 20% Equivalent fuel assemblies 126 Center-to-center spacing of assemblies, in. 21 Maximum keff possible with unborated water 0.90 Spent Fuel Storage Pit (conunon to both units) Core storage capacity 2-2/3 + 10% Equivalent fuel assemblies 464 Number of space accommodations for failed 19 fuel cans 6 Number of space accommodations for spent fuel shipping casks 1 Center-to-center spacing of assemblies, in. 21 Maximum keff possible with unborated water 0.90 Miscellaneous Details Width of refueling canal, ft. 3 Wall thickness for spent fuel storage pit, ft. 3 to 6 Weight of fuel assembly with control rod assembly (dry), lb. 1\.,1570 Capacity of each refueling water storage tank, gal. 375,000 Minimum contents of each refueling water storage tank for Safety Injection and Spray ,e System Operability, gal. Quantity of water required for refueling, gal. 350,000 220,000
9.12.3-1 12-1-69

9.12.3 9.12.3.1 FUEL HANDLING STRUCTURES Refueling Cavities Each reactor cavity is a reinforced concrete structure forming a pool above the reactor when filled with borated water for refueling. The cavity is filled to a depth that limits the radiation at the surface of the water to 50 mr per hour during those brief periods when a fuel assembly

is transferred to the upender and is at the closest approach to the surface of the water.

The reactor vessel flange is sealed to the bottom of the refueling cavity by an inflatable pneumatic seal ring which prevents leakage of refue;l.ing water

from the cavity. This seal is fastened and closed after reactor cqoldown but prior to flooding the cavity for refueling operations. Should the seal deflate during refueling, its passive se~ling design will preclude failure and leakage. During reactor operation, the seal is removed and stored inside the containment structure.

The cavity is large enough to provide storage space for the reactor upper internals, the control rod assembly drive shafts, miscellaneous refueling tools and the lower internals. The walls and floor of the refueling cavity are lined with 1/4 inch type 304 stainless steel .
9.12.3-2 12-1-69 9.12.3.2 Fuel Transfer Canals and Transfer Tubes In each unit a fuel transfer canal extends along one wall of the refueling cavity to the inside surface of the reactor containment. The canal is formed by two concrete shielding walls, which extend upward to the same elevation as the refueling cavity. The floor of the canal is at a lower elevation than the refueling cavity to provide the greater depth required for the Fuel Transfer System upending device and the control rod assembly change fixture located in a slot adjacent to the canal. Each transfer tube provides a connecting passage between the end of the transfer canal in the reactor containment and the spent fuel pit. A gate valve provides a positive closure on the spent fuel pit end of the tube; back-up protection is provided by a blind flange on the reactor cavity end of the tube. Walls and floors of the canals are lined with 1/4 inch type 304 stainless steel. 9.12~3.3 Spent Fuel Pit The spent fuel pit is designed for the underwater storage of spent fuel assemblies and control rod assemblies after their removal from the reactor. It is designed to acconnnodate a total of approximately 464 fuel assemblies, which is about two and two-thirds cores plus 10% of a core, and a shipping cask

9.12.3-3 12-1-69 The spent fuel pit is constructed of reinforced concrete. The entire interior of the pit is lined with 1/4 inch type 304 stainless steel.

Storage racks erected on the pit floor are provided to hold the spent fuel assemblies. Fuel assemblies are placed in vertical cells, grouped in parallel rows having a minimum center to center spacing of 21 inches in both directions. The racks are designed so that it is impossible to insert fuel assemblies in other than the permitted locations, thereby assuring the necessary spacing between assemblies to prevent criticality even if the pit were inadvertently filled with unborated water. Control rod assemblies are stored in the fuel assemblies. Radiation monitors for this area are provided as described in Section

11. 3. 4.

9.12.3.4 New Fuel Storage New fuel assemblies and control rod assemblies are. stored in a separate area of the fuel building where they are unloaded from trucks. This storage area is desi~ned to hold 126 new fuel assemblies in vertical racks and is used nrimarily for the storage of the one-third replacement core plus 10% for each of the two units. The storage of 126 new fuel assemblies is approximately 75% of a complete core. The assemblies which make up the remaining approximately 25% of the first core for each unit are stored in the spent fuel pit. The spent fuel pit is used for the storing of new fuel assemblies only for the initial core loading of of each unit. The new fuel assemblies are stored in racks in parallel rows having a minimum center-to-center distance of 21 inches.
9.12.4...;.1 12-1-69

9.12.4 9.12.4.1

;REFUELING EQUIPMENT Reactor Vessel Stud Tensioners S_tud tensioners :are used *to make up the reactor vessel head closure joint. .During this process all s,tuds are stressed suf.ficiently to hold the *closure .heads .seated and .maintain leak.tightness during operation.

The stud tensioner is a.hydraulically operated device ,provided to permit p:t:eloading .and unloading ,of the reactor vessel.closure studs at cold shutdown :condi.tions.., Stud :tensioners minimize the ,t'ime required for

*the tensienin,g or unloading operations., minimi:ze thread damage., and

permit precision .stud tensioning .. Three tensioners .a:r,e *provided *for

.each 'Unit .and :they are applied simultaneously .to three studs 120° apart.

One *hydraulic pumpin,g uni*t operates the *tensioners *which ,are hydraulically connec*ted .in ,parailel. The studs are tensioned to their operational load in two :Steps to. prevent high stresses in the f:lange region and unequal loadlL'ngs in the studs. 'Relief valves are provi:ded on each tensioner to prevent .ov;e:ttensioning *of the studs due to excessive pressure. Charts indicating the s*tud elongation and load for a given oi:l pressure are included in .the 'tensioner operating instructi*ons. In ,addition, micrometer_s are provided .to measure the elongation of the studs after *1tensioning

9.12.4-2 12-1-69

~.12.4.2 Reactor Vessel He~d Lifting Device The reactor vessel head lifting device consists of a welded and bolted structural steel frame with suitable rigging to enable the reactor containment crane operator to lift the head and store it during refueling operations. The lifting device remains permanently attached to the reactor vessel head. 9.12.4.3 Reactor Internals Lifting Device The reactor internals lifting device is a structural frame suspended from the reactor containment .polar crane. One lifting device is provided for each unit. The frame is lowered onto the guide tube support plate of the internals and manually bolted to the support plate by three bolts with long torque tubes extending up to an operating platform on the lifting device. Bushings on the frame engage guide studs in the vessel flange to provide close guidance during removal and replacement of the internals package. Manipulator Crane The manipulator crane is a rectilinear bridge and trolley crane with a vertical mast extending down into the reactor cavity water. A manipulator crane is provided for each untt. The bridge spans the reactor cavity and runs on rails set into the floor along the edge of the reactor' cavity. The bridge and trolley motions are used to position the vertical mast over a fuel assembly in the core. **
9.12.4-J 12-1-69

A long tube with a pneumatic gripper on the end is lowered down from the mast to grip the fuel assembly. The gripper tube is a telescopic de:v:i,ce that is long enough so the upper end is still contained in the mast when the gripper end contacts the fuel. A winch mounted on the trolley raises the gripper tube and fuel assembly up into the mast tube., The fuel, while inside the mast tube, is transported to its new position.

All controls for the manipulator crane are mounted on a console located on the bridge. The bridge is positioned on a coordinate system laid out on one rail. The Electrical Readout System on the console indicates the position of the bridge. With the aid of a scale the trolley is positioned on the bridge structure. The scale is read directly by the operator at the console. The drives for the bridge, trolley and winch are variable speed, including a separate inching control on the winch. Electrical interlocks and limit switches on the bridge and trolley drives protect the equipment. In an emergency, the bridge, troll~y and winch can be operated manually using handwheels on the motor shafts. The suspended weight on the gripper tool is monitored by an electrical. load cell indicator mounted on the control console. A load in excess of 110 peic~nt of a fuel assembly weight stops the winch drive from moving in the up direction. The gripper is interlocked through a weight sensing device and also a mecha~ical spring lock so that it cannot be opened wheri supporting a fuel assembly .
9.12.4-4 12-1-69 In addition to the travel limit switches on the bridge and trolley drives, the following safety features are incorporated in the system:

1. Bridge, trolley, and winch drives are mutually interlocked to prevent simultaneous operation of any two drives.

2. Bridge and trolley main motor drive operation is prevented except when the GRIPPER TUBE UP position switch is actuated.

3. A solenoid valve in the air line to the gripper is de-energized except when zero suspended weight is indicated by a force gage.

As backup protection for this interlock, the mechanical weight actuated lock in the gripper prevents operation of the gripper under load even if air pressure is applied to the operating cylinder.

4. Hoist drive circuit in the up direction is opened when the EXCESSIVE SUSPENDED WEIGHT switch is actuated.

5. Hoist drive circuit in the up direction is operable only when either the OPEN or CLOSED indicating switch on the gripper is actuated.

6. Bridge and trolley drives are interlocked in the direction of the~

transfer system so that the bridge is prevented from traveling beyond the core area unless the trolley is aligned with the

refueling canal cent~rline. The trolley drive is locked out when the bridge is moved beyond the edge of the core.

Suitable restraints are provided between the -bridge and trolley structures and their respective rails to prevent. derailing due to the Design Basis Earthquake.. The manipulator crane is designed to prevent disengagement of a fuel assembly from the gripper under the Design Basis Earthquake. The manipulator crane is parked to one side of the reactor and secured when not in_

use. The manipulator crane is designed as a Class I component (section 15 *. 2.l~. 9.12.4.5 Motor Driven Platform and Hoist The movable platform with hoists in the fuel building is a wheel-mounted, motor driven platform with overhead trusses supporting electric monorail hois,ts for lifting .new and spent fuel assemblies. The platform spans the spent fuel pit and may be_ maneuvered over any part -of the fuel .building area. The hoist travel and the length of the long £uel handling tool are designed to limit the maximum li_ft of a spent fuel assembly :to assure an adequate water shield above the fuel. .The movable platf_orm is designed as a Class I component and is parked to .one side of the spent fuel racks and secured when not in use. Suitable restraints are provided between the bridge and the rails to prevent derailing during the Design B.asis Earthquake.* 9.12.4.6 Fuel Transfer System The Fuel Transfer System for each unit, shown in Figure 9.12-1, is an underwa_ter conveyor car and trac~stem which extends from the refueling canal through the tra,nsfer tube and into~e spent fuel pit. 'the conveyor
9.12.4-6 12-1-69 car receives a fuel assembly in the vertical position from the manipulator crane after which the fuel assembly is tilted to a horizontal position and passed through the transfer tube to the spent fuel pit. Inside the spent fuel pit it is tilted to a vertical position in preparation for placement in the storage racks, A crane over the spent fuel pit is used for moving the spent fuel and the spent fuel casks (see Figure 15.1-5). During reactor power operation the conveyor car is stored in the containment and a blind flange is bolted on the transfer tube to seal the rea,ctor contain-ment penetration. 9.12.4.7 Fuel Elevator

The fuel elevator lowers new fuel assemblies from the top to the bottom of the spent fuel pit so that the new fuel handling tool and the hook and cable of the traveling platform hoist do not become contaminated by immersion in the pit water. Transfer.of the fuel assembly from the elevator at the bottom of the pit to the upending frame is accomplished with the long fuel handling tool *which also is used for transferring spent fuel. The fuel elevator is a Class I component.

9.12.4.8 Control Rod Assembly Changing Fixture A fixture is mounted on a wall of each reactor cavity for removing control rod assemblies from spent fuel assemblies and inserting them into new fuel assemblies. The fixture consists of two main components; a guide tube mounted to the wall for containing and:guiding the control rod assembly, and a wheel-mounted carriage for holding the fuel assemblies and positioning fuel assemblies under

the guide tubes. The guide tube contains a pneumatic gripper on a wlnch that grip~ the control rod assembly and lifts it out of the fuel assembly.

By positioning the carriage, a new fuel assembly is brought under the guide tube and the gripper lowers the control rod assembly into place. The manipulator crane loads and removes the fuel as-semblies into and out of the carriage. 9.12.4.9 Refueling Water Storage Tank The refueling water storage tank of each unit, described in detail in Section 6.2.2.2, provides the water for filling the reactor cavity and ,:or certain saf.eguards sys terns

9.12.4.10 Fuel Cask Trolley

The crane for handling the spent fuel cask is a trolley of 125 tons capacity running on fixed rails. The rails span the east end of the fuel pit in an area where no spent fuel storage racks are installed. The rails pass over the decontamination building and then over the road way. The fuel cask trolley is designed a*s a Class I component.

Restraints are provided to prevent displacement of the trolley from the rails. 9.12.4.11 Polar Crane The overhead crane in the containment is of the polar configuration and is supported on the circular crane wall. The crane has two main hooks with a capacity of 125 tons each with a maximum hook elevation of 53 ft above the

operating floor. The polar crane has access to the entire area within the

9.12.4-8 12-1-69 crane wall. The crane is designed as a Class I component. the crane can be dislodged during an earthquake. No parts of

Restraints _are provided between the trolley and bridge and between the bridge and rails to prevent derailing during a Design Basis Earthquake.

9.12.5-1 12-1-69

9.12.5 9.12.5.1 REFUELING PROCEDURE Design Bases The refueling operation follows a detailed operating procedure which is established to provide a safe, efficient refueling operation. The following significant points are assured by the refueling procedure:

1. The refueling water contains approximately 2,500 ppm boron. The boron concentration together with the control rods is sufficient to keep the core approximately 10% ~k/k subcritical during the refueling operations. The boron concentration is sufficient to maintain the core shutdown if all of the control rods were removed from the core.

2. The water level in the reactor cavity is high enough to keep the radiation levels within acceptable limits when the fuel assemblies are being removed from the core. This water also provides adequate cooling for the fuel assemblies during transfer operations.

3. Fuel handling operations and equipment are designed so that the possibility of fuel mishandling or damage is minimized

9.12.5-2 12-1-69 9.12.5.2 Preparation Sequence

1. The reactor is shut down and cooled to ambient conditions.

2. The control rod assembly drive mechanism missile shield is removed and stored in the containment.

3. Control rod drive assembly mechanism cables and cooling air ducts are disconnected from the mechanisms and stored in the .containment.

4. Reactor vessel head insulation and instrument leads are removed.

5. The reactor vessel head nuts are loosened with the hydraulic tensioners. *

6. The reactor vessel head studs are removed for testing and storage.

7. The reactor cavity drain holes are plugged and the fuel transfer tube flange removed.

8. Checkout of the fuel transfer device and manipulator crane is started.

9. Guide studs are installed in three holes and the remainder of the stud holes plugged.

9 .12.5-3 12-1-69

10 .- The reactor vessel cavity seal ring is placed in position and inflated.

11. Final preparation of underwater lights and tools is made.

Checkout of manipulator crane and Fuel Transfer System is completed.

12. The reactor vessel head is unseated and raised one foot with the reacto'r containment polar crane.

13. The Reactor Cavity and Fuel Transfer System is filled with water to the level of the vessel flange. The water is pumped.

into the reactor cavity by the low head safety injection pumps from the refueling water stor,age I tank t?rou;gh the reactor. vessel.

14. The reactor vessel head is slowly lifted while additional water is pumped into the reactor cavity. The water level and vessel head are raised simultaneously keeping the water level just below the head.

15. The reactor vessel h~ad is removed to its storage pedestal on the bottom floor of th~ reactor containment.

16. The control rod asse~ly ,drive shafts are unlatched

9.12.5-4 12-1-69

17. The reactor vessel internals lifting rig with the reactor vessel flange protector attached is lowered into position by the containment crane and latched to the support plate.

18. The flange protector is unlatched from the lifting rig.

19. The reactor vessel upper internals package is lifted out of the vessel and placed in the underwater storage stand on the floor of the reactor cavity.

20. Removal, insertion, and shifting of' fuel assemblies proceed in accordance with the refueling sequence.

9.12.5.3 Refueling Sequence The refueling sequence is started. with the manipulator crane. The sequence for fuel assemblies ~ithout control rod assemblies is as follows:

1. Selected spent fuel is removed from the core and placed into the Fuel Transfer System for removal to the spent fuel pit.

2. Partially spent fuel is shifted as required within the core.

3. New fuel assemblies are brought in from the spent fuel pit through the Fuel Transfer System and loaded in the outer region of the core.

l I 9.12.s.,,.s 12-1.-69

4. Whenever new fuel is added to.the reactor core, a reciprocal curve of source neutron multiplication is determined and recorded to verify the supcriticality of the core at periodic intervals.

'rhe refueling sequence is modified, for fuel assembliei:; containing control rod *assemblies, as required. If a transfer of the control rod assemblie_~

betwe_en fuel assemblies is required, the fuel assemblies are taken to the cont-rol rpd assembly chan~e :f:ixture to e~change tl:ie control rqd assembly from one fuel assembly to another fuel assembly. 9.12.5.4. Reassembly Sequence

1. The FQ¢1 '.r~ans;Eer System conveyor car is parked and the fuel

.trap.sfe~ t;ube isolation valve closed.

2. The reac,torvessel internals package is picked ~p by the reactor cont:a:J_µ,.,,.

ment polar crane and replaced in t:he reactor vessel. The reactor vessel ;i.nt:ernals'-lifting rig with flange protector is removed_

to stprage pn the bott;om floor of the reactoi;- conta;i.nment:.

3. rhe cont~ol rod assembly drive shafts are relatched to the control rods.

4. The manipulator crane is parked and s~cured.

5. The react;or vessel head is picked up by the react;or containmeqtpolar-crane and positioned over the reactor vessel

9.12.5-6 12-1-69

6. The reactor vessel head is slowly lowered as the water level is lowered. The water'level is lowered by opening a valve at the residual heat removal pump discharge and water is pumped from the reactor cavity into the refueling water storage tank. The normal residual heat removal line is closed while pumping to the tank.

7. When the reactor vessel head is about one foot above the flange, the reactor cavity is drained. When the water in the reactor cavity is slightly below the vessel flange level, the valve to the refueling water storage tank at the residual heat removal pump discharge is closed. The normal residual heat removal operation is restored and most of the remaining water in the reactor cavity is pumped to the storage tank by the reactor cavity purification pumps.

The small amount of water remaining is drained to the containment sump and then pumped to the Liquid Waste Disposal System.

8. The reactor vessel head is seated.

9. The guide studs are removed to their storage rack. The stud hole plugs are removed.

10. The head studs are replaced and retensioned.

11. The fuel transfer tube blind flange is replaced.

12. Electrical leads and cooling air ducts are reconnected to the control rod assembly drive mechanisms.

9.12.5-7 12-1-69

13.

14. Vessel head insulation and instrumentation leads are replaced. A hydrostatic test is performed on the Reactor Coolant System.

15. Control rod assembly drive operation is checked.

16. The control rod assembly drive mechanism missile shield is picked up with the reactor containment crane and replaced.

17. Pre-operational startup tests are performed.

9.12.5.5 Handling of Failed Fuel Assemblies A fuel assembly considered to be leaking is placed in a failed fuel can

located in the spent fuel pit and the car is sealed*to provide an isolated chamber for testing for the presence of fission products.

The failed fuel cans are stainless steel with lids that can be bolted in place remotely. An internal gas space in the lid provides for water expansion and for collection and sampling of fission products gases. Various remotely operable quick-disconnect fittings permit connections of the can to sampling loops for continuous circulation through the can. If sampling confirms the presence of fission products indicative of a cladding failure, the sampling lines are closed off by valves on the can and the encap-sulated fuel assembly is stored in the spent fuel storage pit to await ship-ment. Design of the can complies with shipping cask design requirements so

that the defective fuel can be stored and shipped while sealed in the failed fuel can.

9.12.6-1 12-1-69

9.12.6 SYSTEM DESIGN EVALUATION Underwater transfer of spent fuel provides essential simplici.ty and safety in handling operations. Water is an effective, economic and transparent radiation shield and a reliable cooling meditnn for removal of decay heat.

Basic provisions to ensure the safety of refueling operations include:

. 1. Gamma radiation levels in the containment and fuel storage areas are continuously monitored. These monitors provide an audible alarm at the initiating detector and in the Main Control Room indicating an unsafe condition. Continuous monitoring of reactor neutron flux provides irrnnediate indication in the Main Control Room and alarm of an abnormal core flux level.

2. Violation of containment integrity is not permitted when the reactor vessel head is remove~ unless the shutdown margin is maintained greater than 10% ~k/k.

3. Whenever new fuel is added to the reactor core, the reciprocal curve of source neutron multiplication is monitored to verify the subcriticality of the core.

4. Adequate supervision and planning of operation

9.12.6-2 12-1-69 9.12.6.1 Incident Control Direct communication between the Main Control Room and the reactor davity manipulator crane will be established whenever changes in core geometry or conditions are taking place. This provision allows the control room operator /

to inform the manipulator operator of any impending unsafe condition detected from the Main Control Room indicators during fuel movement. 9.12.6.2 Malfunction Analysis An analysis is presented in Section 14 concerning damage to one complete outer row of fuel elements in an assembly, assumed as a conservative limit for evaluating environmental consequences of a fuel handling incident.
9.12.7-1 12-1-69

9.12.7 MINIMUM OPERATING CONDITIONS Minimum operating conditions for the Fuel Handling System are contained in the Technical Specifications .

9.12.8-1 12-1-69

9.12.8 TESTS AND INSPECTIONS Prior to initial fueling, pre-operational checkouts of the fuel handling equipment are performed to ensure proper performance of the fuel handling equipment and to familiarize operating personnel with operation of the equipment. A dummy fuel assembly is used for this Upon completion of core loading and installation of the reactor vessel head, certain mechanical and electrical tests are performed prior to initial criticality. The electrical wiring for the control rod assembly drive circuits, the control rod assembly position indicators, the reactor trip circuits, the in-core thermocouples, and the reactor vessel head water

temperature thermocouples is tested at the time of installation. The tests are Fepeated on these electrical items before initial operation.

Prior to subsequent refueling operations, the equipment will be inspected for operating condition and certain components, such as the fuel transfer car and manipulator crane, will be operated to ensure reliable performance prior to moving irradiated fuel. Pre-refueling checks are part .of a con-tinuing program

9.13.1-1 12-1-69

9.13 9.13.1 AUXILIARY VENTILATION SYSTEMS GE:tjERAL DESCRIPTION The nucle~_:t::* Auxiliary Ventilation System diagrams are shown in Figure 5. 3-1.

These include the Ventilation and Heating Systems for the auxili:ary building, fuel building, decontamination building, cable vaults, safeguards areas adjacent to the reactor containments ar~d the Main Control and relay roQm area. The auxiliary building, fuel building, decontamination building, Main Control Room and ventilation vent are shared by the two units. Individual cable vaults and safeguards areas, and relay rooms are provided for each unit. The Main Control Room and relay rooms for both units are in the service building *

The auxiliary building is a four-level compartmented structure containing the auxiliary nuclear equipment for both units. Equipment handling radioactive fluids is located on the lower three levels, isolated and shielded as required,,

and the upper level is a ventilation equipment room. Waste gase:s* with a relatively high potential for radioactivity are discharged through filters or the Gaseous Waste Disposal System to the process vent (See Section 11.2.5.1). The ventilation exhausts from some primary plant areas are subject to comparatively slight radioactive contamination from such ,limit_ed sources as pump gland or pipe weepage. The following features are incorporated in these exhaust systems to protect the environment from this relatively remote contamination possibility:
9.13.1-2 12-1-69

1. Two 50 percent capacity exhaust fans installed in parallel, with an automatic back-flow damper on each fan. This provides approximately 70 percent capacity exhaust in the event one fails, and a step flow reduction capacity in the event of radioactive contamination.

2. Sampling lines for selective contamination monitoring from the indi-vidual exhaust ducts to the Main Control Room.

3. A common iodine filter bank of two filter assemblies, each consisting of roughing, absolute, and charcoal filters capable of handling the largest exhaust system. This arrangement provides an effective standby filter if one assembly becomes saturated.

4. Exhaust bypass arrangements for selective filtration of any exhaust system. All bypass and filter dampers are remote-manually operated from the Main Control Room as required. In* addition, the safeguards areas systems are automatically bypassed*upon a containment con-sequence limiting signal to ensure that any possible airborne radio-active leakage will be removed. During refueling, the fuel building exhaust will be continuously bypassed,through the charcoal filters to ensure radioactive removal in case of airborne contamination from any source.

5. Exhaust to the atmosphere through a common, continuously monitored ventilation vent discharging upward at El. 140 ft-6 in., with a velocity in excess of 4,000 fpm. For details of monitoring equipment and diversion of ventilation control, see Section 11.3.3.

9.13.2-1 12-1-69 9.13.2 DESIGN BASIS Outside ambient conditions used for design purposes are 930 F.summer dry bulb, 780 F SUIJ1.IDer wet bulb, 730 F summer dew point, 100 F winter dry bulb, 580 Fall year ground temperature, and 15 mph all year wind velocity. Ventilation is based on limiting the temperature in occupied spaces to 1000 F, limiting the temperature in normally unoccupied machinery spaces to 1200 F, or providing 10 air changers per hour. Provision~ are also made, by dual fans. installed in parallel or two speed fan motors, to reduce air quantities*in mild weather and during the heating s.eason.

Ventilation for huclear auxiliary systems is designed on a once-through basis

Supply ~iris introduced to areas least likely to be contaminated and exhausted directly from those with the greatest contamination potential.

Filter banks for radioactive contamination are designed to remove 99.97 percent of solid particles down to 0.3 micron in size and 99.9 percent of any methyl iodide or iodine. vapor entrained in the ventilation exhaust. Filter assemblies consist of roughing, parti~ulate and charcoal filters in series. The Main Control Room and computer room air conditioning is designed to main-tain 750 F dry bulb and 50 percent relative humidity during either normal or emergency conditions. The relay rooms are designed for 800 F dry bulb and 40 percent relative humidity during normal conditions, and 87° F dry bulb and

35'percent relative humidity during emergency operations.

9.13.2-2 12-1-69 The Main Control and relay room area exhaust and replenishment supply venti-lation is provided by external systems for normal operations. In an emergency the Main Control and relay room area is sealed with weatherstripped doors and tight external duct,closures. The air conditioning systems will continue to operate normally. Bottled compressed dry air is installed to provide fresh breathing air and maintain a positive interior pressure to ensure outward leakage to preyent contamination from the outside. The bottled air will last approximately 1 hr. Emergency supply ventilators;- taking suction from the turbine building th;ough roughing, particulate and iodine filters, are pro-vided to continue the supply of breathing air and pressurization indefinitely upon depletion of the bottled air supply. Air conditioning and associated auxiliary equipment required to operate during emergency conditions can be powered from emergency buses. The ventilation exhaust from the safeguards areas to the ventilation vent, and the ventilation vent, meet Class I design criteria (Section 15.2). Air con- ~itioning and emergency ventila!:ion equipment for the Main Control and relay room area also meet Class I design criteria. Ventilation System Arrangements are shown in Figures 9.13-1, -2, -3, -4.
9.13.,3~1 12-1-69, 9.13.3 SYSTEM DESCRIPTlONS 9.13.3.*1* Auxiliary Building Ventilation The auxiliary building is supplied with air 'by two 50,000 cfm air handling units. The systems have automatic roll filters for continuous cleaning and steam coils for winter heating. Four exhaust fans in subsystems are used: one of 55,000 cfm for the central spaces, one of 43,000 cfm for the remainder of the potentially contaminated spaces, one of 2,500 cfm for the clean resin and cement storage rooms, and one of 2,800 cfm from the elevator machinery room. The exhausts with radioactive contamination potential are each fitted with two 50 percent capacity fans in parallel and always discharge through the ventilation vent.

These exhausts can be diverted remotely through the common filter bank from the Main Control. Room as described in Section 9.13.1. Particulate filters are installed in the exhaust branches from the auxiliary building decontamination room and sample cooler spaces for continuous filtration. The nonradioactive*

exhausts are discharged directly to the atmosphere. Spaces subject to radioactive contamination have exhaust intakes located as far removed from the space access as feasible. The resulting negative pressure draws the makeup air in through the access and sweep~ the space with supply air so that any airborne contamination from equipment leakage will be drawn inward to the exhaust

9.13.3-2 12-1-69 9.13.3.2 Fuel Building Ventilation The ventilation provides heating to 90° F to inhibit the buildup of con-densation, high efficiency filtration to reduce the possibility of clouding the spent fuel pit annan excess exhaust flow to maintain a negative pressure in the building for inward leakage. Two-speed supply fans and dual exhaust fans are provided to permit step capacity reduction in case of airborne contamination and to reduce steam requirements for winter heating.

Two supply fans are provided, one of 29,000 cfm capacity serving the spent fuel pit, and one of 5,000 cfm capacity for the remote equipment space at El. 6 ft-10 in. Both take suction from a common plenum fitted with a combination roll and high efficiency filter (95 percent NBS atmospheric dust) and steam coils ~ for space heating. Heating control, both summer and winter, is as follows: 0

1. 75 F minimum summer and winter inside temperature.

2. Vary the temperature difference between inside and outside from 30° F ~Tat 45° F outside to 15° F 6T at 75° F outside.

0

3. Terminate heating at 90 F inside temperature.

Dual exhaust fans of 17,500 cfm capacity each discharge through the ventilation vent. This exhaust will be continuously diverted through the common iodine filter:bank during refueling. The section of the equipment space subject to radioactive contamination is exhausted directly by a branch duct from the decontamination building exhaust system.
9.13.3-3

9.13.3.3 Decontamination Building Ventilation 10-15-70 The decontamination building is ventilated at approximately 15 air chaxlges per hour and arranged to maintain a negative pressure for inward leakage.

The supply system incorporates a continuous roll filter and steam coils for space heating. Two-speed supply fan control is available for capacity reduction to provide greater inward leakage if required. Dual exhaust fans discharge through the ventilation vent with a filter bank remote-manual bypass arrangement *

9.13.3.4 Safeguards Areas Ventilation The safeguards areas are outside of, and adjacent to, each reactor containment structure. They contain the recirculating spray pumps, low head safety injec-tion pumps, refueling water recirculation pumps, containment spray pumps, auxiliary steam generator feed pumps, motor control center, and cable vault.

The areas with a significant contamination potential (recirculation spray pump and low head safety injection pump and valve spaces) are exhausted by 6,150 cfm capacity dual fans located in the auxiliary building which discharge to the ventilation vent. An automatic capability is provided for particulate and iodine filtration on a higp-high containment pressure signal. The remaining spaces are exhausted directly to the atmosphere. Heated supply air is provided for all spaces except the auxiliary steam generator feed pump area where the
main steam lines provide sufficient space heating. 9.13.3-4 10.,...15-70 The 16,000 cfm intake supply systems are fitted with continuous roll filters and steam heating coils for cold weather space heating. 9.13.3.5 Service Building Ventilation The ventilation for service building spaces subject to possible radioactive contamination is described below. The hot laboratory, count room, and heal th physics lab are exhausted by two 2,300 cfm fans in parallel. The exhaust is continuously drawn through roughing and particulate filters and discharged through the monitored ventilation vent

The laundry and adjacent area subject to possible contamination from soiled clothing is exhau~ted by a 4,000 cfm fan. The effluent is continuously drawn through a.4 cell roughing and particulate filter bank and exhausted through the ventilation vent.

A special machine shop local exhaust system of 1,600 cfm is installed for use when contaminated equipment is being worked on. Portable flexible ducts are provided for taking suction directly from the work area. The system exhausts through a 2 cell roughing and particulate filter assembly to the monitored ventilation vent. Ventilation exhausts for the remainder of the service, turbine and yard buildings are discharged directly to the atmosphere.
9.13.3-5 10-15-70 9.13.3.6 Main Control and Relay Room Area Ventilation All of the air conditioning equipment for the Main Control and relay room area is located within a tornado and missile protected structure to ensure cooling at all times during both normal and accident conditions, The Main Control and re.lay room area is air- conditioned by independent 100 percent redundant air handling units installed within the space served. One unit for each space is supplied by a separate chilled water supply, with a second supply serving the alternate unit; thus, complete redundancy is provided in case of malfunction of either a unit or its chilled water supply. Three 100 percent refrigeration chillers of 90 tons each serve these two groups to

ensure that two are operable whenever one is dOWH for maintenance. Condensing cooling water is provided by two independent gravity supplied service water lines. The service water system is described in Section 9.9.1.

Main Control and relay room area exhaust and makeup air of 2,500 cfm is provided by other systems. Tight remote-manually operated closures in these ducts, and weatherstripped doors, permit pressurization of the Main Control Room area during an accident with bottled compressed dry air. The bottled air bank provides 18,000 scf of free air, sufficient for approximately an hour's pressurization. Emergency ventilation is provided for each space. These take suction fro~ the I turbine building through roughing, particulate and iodine filters to remove any airborne radioactivity. They will indefinitely extend Main Control and relay room area pressurization and the supply of breathing air upon depletion
of the bottled compressed air. 9.13.3-6 12~1-69 Control of the air* conditioned and pressuriza-tion systems is manual-remote from the Main Control Room. Emergency power is supplied for all air conditioning and emergency ventilation equipment (Section 8.5).

portable fire extinguishers are located in each room of the Main Control and relay room area. The Main Control Room is separated from the relay rooms by a 3 hr fire resistant barrier. The relay rooms are separated by a 2 hr fire wall. Thus, fire or smoke in any space can be isolated.

TABLE 9. 13-1 MAIN CONTROL AND RELAY ROOM AREA VENTILATION AND AIR CONDITIONING SYSTEMS - DESIGN DATA Number Unit Refrigeration of Capacity Static Pressure Motor Capacity Filter Service Units Cfm in. W.G. Hp MBh Type Main Control Room Air Conditioning 4 10,000 and 2 10 234 Rough 12,000 Unit 1 Relay Room Air Conditioning 2 10,000 2 10 355 Rough Unit 2 Relay Room Air Conditioning 2 10,000 2 10 355 Rough Control and Relay Room Air Supply 2,500 (Branch from Assembly Room Unit) Roll Control and Relay Room Area Exhaust 1 2,500 0.75 1 Main Control Room Emerg Ventilation 2 1,000 4 1.5 0 Charcoal Unit 1 Relay Room Emerg Ventilation 1 1,000 4 1.5 0 Charcoal Unit 2 Relay Room Emerg Ventilation 1 1,000 4 1.5 0 Charcoal Control and Relay Room Area Compressed Dry Air System 1 18,000 cu ft free air

..... I.O N

I I-'

1-'W I

O'I w I.O I

-...J

9.13.4-1 12-,-1-69

9.13.4 DESIGN EVALUATION The ventilation systems in areas of potential contamination provide contamina-tion control by ensuring that air is not recirculated, 10 or more changes per hour are supplied, and the air is supplied to the least likely areas to be contaminated for circulation to locations subject to the greatest contamination potential for exhaust. After being monitored for gaseous and particulate activity, the systems are exhausted through a stack 114 ft above grade dis-charging upward at a velocity in excess of 4,000 fpm._ A capability is provided for all nuclear Auxiliary Exhaust Systems subject to* airborne radioactive con-tamination to be bypassed through roughing, particulate and acUvated charcoal filters *

The Ventilation System limits summer space temperatures to 100 spaces and 120 0

Fin normally unoccupied machinery spaces. 0 Fin occupied Ventilation is based on the heat produc:f,.ng equipment operating, _and summer space temperatures will be lower whenever such equipment is down for maintenance. The heating system provides space temperatures sufficient for winter operations and/or the inhibition of condensation in the fuel building and below grade spaces. The Main Control and relay room area is completely enclosed in a tornado and missileproof concrete structure that requires continuous air conditioning to keep them operating reliably. Two 100 percent independent air conditioning circuits, each served by one of three 100 percent water chillers, provide the
9.13.4-2 12-1-69 essential redundancy. These systems are powered from the emergency buses during loss of power and gravity flow service water provides condensing capability under all conditions. 9.13.4.1 Incident Control

The safeguards area exhau~t to the ventilation vent is automatically bypassed through the particulate and iodine filters upon a containment consequence limiting accident signal. Inward leakage is ensured as the exhaust and supply are balanced to produce a negative pressure. In the event of leakage from the Recirculation Spray System, airborne radioactivity would be removed from the safeguards area by these filters.

During refueling, the fuel building exhaust will be continuously diverted through the particulate and activated charcoal filter banks. This will remove any airborne radioactivity~ On a high radiation alarm from the ventilation vent continuous monitor the Main Control Room operator will:

1. Trip the supply fans and one exhaust fan for:

Auxiliary Building Central Area

Auxiliary Building General Area Fuel Building Decontamination Building

9.13.4-3 12-1-69

2. Bypass the remaining exhausts through the particulate and charcoal filter banks. This will remove the contamination,and avoid dis-charging radioactivity to the environment.

3. Sample the two safeguards exhausts and then the above exhaust systems in turn. The safeguards areas are not included in the above alignment as their only source of contamination is the containment recirculation spray, and this exhaust is automatically bypassed whenever the spray is operating. However, the safeguards areas will be sampled first as a precaution. As an area is eliminated as the radioactive source, it will be realigned to its normal mode. The contaminated exhaust will continue to be discharged through the filters or shut down as

required until the radioactivity source is eliminated.

In the event of a loss-of-coolant accident, the Main Control and relay room area is sealed off by closing the weatherstripped access doors and the pressure tight external duct closures at the space boundaries and internal fire barriers. The air conditioning will continue to operate normally with-out change. Out flow of air from the Main Control Room will be maintained by area pressurization to approximately 0.25 in. W.G. by breathing quality air introduced from compressed air bottles. Upon depletion of the bottled air, in approximately an hour, the emergency ventilation will be energized to continue pressurization as a precaution. The emergency ventilation is filtered through roughening particulate and iodine filters to breathing quality, and the quantity is sufficient to permit opening the exhaust duct for natural discharge of stale air. All functions are manually controlled from the Main Control Room ventilation control board.
9.13.4-4 12-1-69 Incipient fires in the Main Control and relay room area will be extinguished with portable equipment. If a fire becomes uncon~rol~able, the affected space will be isolated by closing the fire doors. The air conditioning ductwork is sei.f-contained within each space, and the closures in the replenishment air and exhaust ducts fitted at each fire barrier will prevent smoke contamination in adjacent spaces. If the Main Control Room becomes untenable because of fire or smoke, the reactor units can be shut down to a hot standby from their respective auxiliary control areas in the relay rooms. 9.13.4.2 Malfunction Analysis If one of the two safeguards area or fuel building exhaust fans become inopera-tive, the supply fan.will be run at half speed to ensure inward leakage. One exhaust fan provides approxima~ely 70 percent of the total cap~city. The two particulate and iodine filter banks provide 100 percent redundancy. If the Main Control and relay room area air conditioning system should fail, the second 100 percent unit is energized from the Main Control Room. The Emergency Ventilation System will be started immediately if the bottled com-pressed air fails. Each Emergency Ventilation System is sufficient for pressurization of the Main Control and relay room area. 9013.4.3 Tests and Inspections The systems are inspected, tested and balanced upon installation and tested periodically thereafter. Operating hours will be equalized on redundant
9.13.4-5 12-1-69

systems, such as the Main Control and relay room area air conditioning.

Particulate and charcoal filters are individually tested by the manufacturer after fabrication and again after installation. Replacement filters will be tested in the same manner. Filter banks can be tested for leakage and dioctylphthalate smoke test efficiency while in place, and defective cells identified for removal and replacement observing the applicable maintenance procedures. Equipment installed for emergency use is tested during installa-tion and operated monthly thereafter to ensure ?roper functioning

FIG. 9.13-1.

OCT. 15, 1970 - EXH F~N 1-vs-F-S.~ 21100 CFl't

- ELEV Mb.C.\-1 ~OOM ON ,.Q *\.*'2.'l!l" * ('TY .. I CAI..)---*

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I PLA.IJ 1:-L 2 7'- 7 Di DD NOT*'II: tJ0 .*CAL.lfr

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(0000) WINTER All\ QUANTITY L£GEND1 stw STD*Mll*SA,58,SC, fSO SAO* SELF ACTING D~MPll\ .\ PART PLAN £:-L 4"7~4-" VENTILATION ARRANGEMENT PRIMARY PLANT SYSTEMS SHEET I SURRY POWER STATION
K FIG. 9.1:3-2 MAR. 15, 1971, FROM DECJNTAMINA.TlON BLDG

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,;*[Z~;~~Q~#~,s:;;:,=l==::,z~~========::=;~==~~u~>~ PRIMARY PLANT SYSTEMS FROM-+,*=-==~'--..U...t--=~=~~ SURRY POWER STATION BORON * '

R.E:COVERY DL,',.N EL 2'-0" P\J~p HSE PLAN ABO\IE )

"FIG. 9.13-3

H OCT.15,1970 8 ~ g "~

0 "' W

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~ " (FB*Z B)

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FIG. 9.13-4 OCT.15, 1970 I A B C D

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9.14-1 12-1-69

9.14 DECONTAMINATION FACILITY The decontamination facility, see Figure 9.14-1, is a poured concrete and concrete block structure on the north side of the fuel building under the fuel cask trolley rails. This location makes it accessible for tr~nsporting major items to be decontaminated in and out of the building. Roof hatches and a rolling steel door provide access for equipment

9.14.1-1 12-1-69

9.14.1 DESIGN BASES The facility is designed to provide an area in which equipment can be decontaminated without releasing activity to the environment in an uncontrolled manner. Decontamination procedures are specified to reduce surface contamination to a level such that the components can be handled in a safe manner .

9.14.2-1 12-1-69 9.

14.2 DESCRIPTION

The decontamination building is a poured concrete and concrete block building abutting the east end of the fuel building's north wall. A 125 ton trolley runs through a high bay portion of the building immediately adjacent to the fuel building, and over the roof of the remainder of the decontamination building. Three roof hatches permit lowering casks or other objects from the*trolley into the building. A tramrail in the building permits the movement of small parts between work areas and tanks with minimwn personnel exposure. AT-shaped rolling steel door encloses the high bay area from the outside when the trolley is not in use. The fuel building and decontamina-a weathertight tion building are separated by structural gap to permit independent motion of the buildings in the event of an earthquake. Ventilation air is exhausted from the decontamination building through the monitored Ventilation Vent System. On a high alarm by ventilation vent monitors, the decontamination exhaust is remote manually diverted through charcoal filters as described in Section 9.13. The exhaust capacity is greater than the supply capacity of this system, thus producing a slightly negative pressure in the buildings so that all air leakage is inwards.

H*.,

Personnel normally enter the decontamination building by passing through the fuel building from the auxiliary building. An emergency exit from the decontamination building is provided. Liquid wastes from decontamination
9.14.2-2 12-1-69 work are piped to the Liquid Waste Disposal System (Section 11.2.3) for processing. The interior surfaces of the building are covered with suitable materials to permit easy decontamination. A stainless steel pad is provided to protect the floor under heavy objects. Hose connections are provided for compressed air and primary grade water at each work area. An ultrasonic cleaning tank is provided for immersion cleaning of small

\

components. This equipment permits decontamination of equipment by wash down combined with scrubbing, soaking with or without agitation. The various decontamination methods provide a flexibility that wiil give the best decontamination for a specific job, minimize personnel exposure, and limit the release of radioactive material to the environment. Technical information on the equipment provided in the facility is given in Table 9.14-1. Working personnel will be monitored by Health Physics monitors to insure that established exposure* limits are not exceeded. A contaminated solution holdup tank is provided to receive spillage from equipment, r~noff from cleaning operations, and disposal of cleaning solutions. This tank has a pump for transferring liquid to the Liquid Waste Disposal System. A filter is provided for preliminary cleanup of the fluid prior to pumping out.
9.14.2-3 12-1--69

TABLE 9.14-1 DECONTAMINATION FACILITY COMPONENT DATA Contaminated Solution Hold-Up Tank Number 1 Capacity, gal 2,000 Design pressure, psig 30 Design temperature, °F 150 Operating pressure Atmospheric Operating temperature Ambient Material Stainless steel Type 316L Design Code ASME VIII Contaminated Solution Hold-Up Tank Filter Number 1 Retention size, microns 10 - 25 Filter element material Synthetic fibre or metallic Capacity normal, gpm 20 Capacity maximum, gpm 20 Housing material Stainless steel Type 316L Design pressure, psig 150 Design temperature, op 150 Design Code ASME VIII

9 .14 .2-4. TABLE 9.14-1 (Cont'd) 12-:1-69 Contamination Solution Hold-Up Tank Pump Number 1 Type Centrifugal Motor horsepower, H.P. 1 Seal type Mechanical Capacity, gpm 20 Head at rated capacity, ft 50 Design pressure, psig 100 Materials

*Pump casing Stainless steel Type 316 Shaft Impeller Stainless steel Type 316 Stainless steel Type 316

9.14.3-1 12-1-69

9.14.3 DESIGN EVALUATION The facility provides a contained area with all discharges controlled to prevent the inadvertent release of activity to the environment.

9.14.3.1 Malfunction Analysis In the event of leakage from piping or equipment, all areas of the building are provided with sumps to which fluids will drain. The sumps discharge to the Liquid Waste Disposal System. Airborne particulate matter is retained within the building because of the slightly subatmospheric pressure and is discharged in a controlled manner through the monitored

Ventilation Vent .

9..14.4-1 12-1-69

9.14.4 TESTS AND INSPECTIONS Periodic tests are conducted on the radiation detection equipment in the Ventilation System.

Operating equipment and storage tanks are subjected to periodic visual inspections .
FIG.9.14-1. C H K L OCT.15,1970

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10-i ..,. TABLE OF CONTENTS 12-1-69 Section Title Page 10 STEAM AND POWER CONVERSION 10.1-1 10.1 GENERAL DESCRIPTION 10.1-1 10.2 DESIGN BASES 10.2-1 10.3 SYSTEM DESIGN AND OPERATION 10.3.1-1 10.3.1 MAIN STEAM SYSTEM 10.3.1-1 10.3.1.1 Design Basis 10.3.1-1 10.3.1.2 Description 10.3.1-2 10.3.1.3 Performance Analysis 10.3.1-7 10.3.2 AUXILIARY STEAM SYSTEM 10.3.2-1 10.3.2.1 Design Basis 10.3.2-1

10.3.2.2 10.3.2.3

10. 3.3 Description Performance Analysis TURBINE GENERATOR 10.3.2-2 10.3.2-3 10.3.3-1 I

10.3.4 CIRCULATING WATER SYSTEM 10.3.4-1 10.3.4 .. 1 Design Basis 10.3.4-1 10.3.4.2 Description 10.3.4-2

10. 3 .4. 3 Performance Analysis 10. 3 .4-6

10.3.4.4 Tests and Inspections 10.3.4-7 10.3.5 CONDENSATE AND FEEDWATER SYSTEM 10.3.5-1 10.3.5.1 Design Basis 10.3.5-1 10.3.5.2 Description 10.3.5-3 10.3.5.3 Design Evaluation 10.3.5-5 10.3.5.4 Tests and Inspections 10.3.5-6 10.3.6 CONDENSER 10.3.6-1

~. 10.3.6.1 Design Basis 10.3.6-1
10-ii i2-1-69 Section

10. 3.6. 2 10.3.6.3 Title Description Design Evaluation Page 10.3.6-1 10.3.6-3 10.3.6.4 Tests and Inspections 10.3.6-3 10.3.7 LUBRICATING OIL SYSTEM 10.3.7-1 10.3.7.1 Design Basis 10.3.7-1 10.3.7.2 Description 10.3.7-1 10.3.7.3 Design Evaluation 10.3.7-2 10.3.7.4 Tests and Inspections 10.3.7-2 10.3.8 SECONDARY VENT AND DRAIN SYSTEMS 10.3.8-1 10.3.8.1 Design Basis 10.3.8-1 10.3.8.2 Description 10.3.8-1 10.3.8.3 Performance Analisis 10.3.8-3 I 10.3.8.4 10.3.9 10.3.9.1 Inspections and Testing BEARING COOLING WATER SYSTEM Design Basis 10.3.8-3 10.3.9-1 10.3.9-1 10.3.9.2 Description 10.3.9-2 10.3.9.3 Perfor9ance AnalJsis 10.3.9-2

10.1-1 12-1-69 10 STEAM AND POWER CONVERSION 10.1 GENERAL DESCRIPTION This section describes that category of systems and equipment which are required to convert steam energy to electrical energy. The following sections describe separate equipment and systems required for each unit:

1. Main Steam System

2. Turbine Generator

3. Condensate and Feedwater System

4. Condenser

5. Lubricating Oil System

6. Bearing Cooling Water System

7. Auxiliary Steam System The following sections describe those systems which are shared in the operation of both llllits:

1. Circulating Water System

2. Secondary Vent and Drain System

10.2-1 12-1-69 10.2 DESIGN BASES The design bases of the Steam and Power Conversion equipment and systems is largely derived from past design experience with fossil fueled stations and has evolved over a long period of time. Specifically, the design bases are oriented to a high degree of operational reliability at optimal thermal performance. The performance of the collective equipment and systems is a function of environ-mental conditions and the selection of design options. Therefore, the principle design basis is represented by the design heat balances which incorporate all of the applicable design considerations. The heat balances are shown in Figures 10.1-1 and 10.1-2. Figure 10.1-1 shows the heat balance for the extended rating equivalent to 2,546 MWt. Figure 10.1-2 shows the heat balance for the rated operating condi-tion and is applicable to the 2,441 MWt unit rating. The conventional design bases have been modified in order to provide suitability for nuclear application, and these include provisions for specific earthquake, tornado, and missile protection as further described in other sections. Turbine Building Figures 10.1-3 through 10.1-10 show equipment locations

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LB PCIWER FLOM. pomms FEfl HOUR ELEVATION - APPFHJXIHATELT 27 FEET RBtlVE SEA LEVEL. EQUJPHENT. THE REQUJRENENTS OF THE fCILUJIUNCi HAJOR 2862 l\ll6.117 H ENTHALPY, BTU PER POUND llS.00 0.93 1.7ij 5,686,961 980.811 999.35 822,871.i 786,906 10211.i.26 1010l&. 72 10566.59 EQUIPHENT VART WITH LCIAO. 1239 11ll6. l,l.7 F TEHPERRTURE, DEGREES F STEAH GENERATOR BLOHDOHN - 0.10 PERCENT Of STEAH GENERRTlON. 55.00 l.16 1. 75 5,727,350 960.96 999.51.1 823.1&76 767,511 10206. 70 10097.33 10558.t.17 1622 1 ll&6.l,l.7 TO TERHJNAL Ill fFERENCE 2 OPEAAl JNG 1Jlf,lj2 1198. 30 DC OllAl N COOLER APPRCIACH PRESSURE DROPS - PRESSURE ORCIPS FOR THE STERN GENERATORS, 3 Ci:"JNOENSRTE PUHPS 6S.OO 1.119 I. 76 5,775,5511 981. 11 999. ?7 8211.192 788,230 10197. 711 10098. 56 105118.811 REHEATERS, HIHSTURE SEPARATORS, HAIN STEAH, H.P. EX- 2 H. P. HEATER DRAIN PUHP5 CIPEflAT I NG H 0 0.0 KW KILOWATTS 75.00 1.96 1. 96 5,829,301 965.36 1000. qq 8211.293 788,335 10196.1&8 10087.'33 1051,17.115 JN HG PRESSURE, IN OF HERCURT, ABS. HAUST, REHEAT, ANO EXTllACTION PIPING WERE CALCULATED 2 L. P. HEATER ORRIN PUt1F5 CIPEAAl I NG AT All LOADS, 2 STEAH GENEflATDfl FEED PUNPS QPEAAT lNG 85.00 2.59 2.59 5,867.6111 997. 71 1006. 01 616.130 780,176 10299.60 l01BB. 211 10657.76 PSIA PRESSURE, LB PEIi SQUARE IN, RBS. 765,057 10668,38 181 THROTTLE OR INTERCEPT VALVE PRESSURES - TURBINE FLANGE PRESSURES ARE SHOWN ON EXTRAC- ~~- 3.113 3.4.'3 5,91.18,918 1009.97 1015. 55 801,006 101196.211 10380.61 TION LINES ADJACENT TO THE TURBINE. HERTEfl JNLET PRESSURES RAE SHOWN ADJACENT TCI THE HEATERS. THE FflllQWJNG ARE lHPflRTRNT AUXILIARY PQHEA lTEHS WHICH RllE CONSIDERED NOT TO VRRT lo!ITH LORD. J:k1 CONTROL VAL V"" AUXILIRRT POI-IEA FOR THESE RNO LESSER EQUJPHENT HRS BEEN BASE BALANCE CONDITIONS - THE HEAT BALANCE Ct'.INOJTICINS SHt'.l~N CALCULRTEO FOR AVERAGE OAILT AEIJUlAEHENTS. CIN THE DIAGAAH CQRAESPOND TO A ClflCUL.ATJNG MRTER TEMP-ERATURE OF 75 F. CIRCULATING WATER PUHPS PAI HART COQLRNT PUHPS TR8ULATEO CONDJTiaNS - ALL. VALUES SHOWN IN THE TABLES MERE BEAAJNG Cl!l:lllNG WRTER PUHPS DEVELOPED BY CALCULATING CCINPLETE HEAT BALANCES AT A COHPOt.lENT COOLING WATEfl PUHPS FJXED lHACITTLE FUIW AND VARYING THE CIRCULATING WRTER INSTRUMENT AIR' COHPAESSOA TEHPERRTURE, NET TURBINE HEAT RATE" GENERATICIN - STE~~A~~~~R~~~~ FEED PUNP POWER "'e2:::~:-~16,~~~q = 101S6.l,l8 BTU/KW-HR ~* ALL HEAT BALANCE RESULTS ARE FRODlJCED B'I' A COHFUTER PAOGRRH PROCESSED CIN RN IBH 360 l'IOOEl 65 CONPUTER. BALANCE OlAGARl'IS ARE llENERATEIJ BT A CALlFrlRNlR COMPUTER PROOlJCTS INC. FEN PLOTTIHG STSTEH HEAT BALANCE DIAGRAM MAXIMUM GUARANTEED LOAD SURRY POWER STATION
FIG.10.1-3 OCT. 15, 1970

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,7 UHIT ... ' ~ - ~ ,... n V r;':i~Tr \./,TH PTIITR- I _

llELAY ROOM I

DA.A.IN C.OOLEliZS. I rn'-- FU,IP \*SW-P-46 II EL9'-~' ttI 11 ;i, :: . I I tl. *,

~Clf,DC*li!. 1-C.IHIC*IA.

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Et{TIL....TION VENT E, E-E PLE.NUM

* (FM*""' ..8 4 ~C) ~UNIT SUPPL't ~AN &WITCHGE.AR ROOM EL sa'E.'

CABLE VAULT

!JOTES*,

VOLTAGE Rt(,ULATOR

~CALE.._ >'&~ .... ,,_o*

GEN,..,OTE~ ~ IO.EF, l)W,i,.~ F"M - loA CONTROL ROOM EL 21*-o* RELAY ROOM 0 l1'1DRAl.lME PIJW) TURBINE BUILDING F-F SHEET-6 (CM

G,A ... B u,c)

SURRY POWER STATION
FIG. 10.1-9

.. I . I r II y I " I ' I

I

" I OCT. 15, 1970 0 9-V'U-9 '1717 I I I

l I I I I

. -~--ii I I I l

I I

- i -

lCHORD BO"TTOM cu: 1~U":i'5ES EL 113*.9~

; .. l l i - -------f----- - 11-1-----,--+-yVI ,i il fl! r -ff TOP or CRANE RAll.S EL ~Efra~

1-H--'-,C~E CAB r l LP EXHAUST NO. I

,,,,- i-,...

V-~ l.P EXHAUST NO. 2.

..-r-...

i1~ I 11~t1.-MYDROGEN

WITHOI\AWAL COOLER

- ~~~~--~~~~~~;~I = ~~=~~~*~ . ,I lr *. -j--l:::1J>----'---1" I .:

r1-H D /

-H I tMOISTURE.1 f1 \ ,i I(;

i *,ii*, i 1

, i,',i: ,-+--L...--llLJ,l,)----,.r;,c:r.-,lC,,m I

r*~GENERA.TOR.

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ll ~~T--..L.~.,.:::::::::c .. µ./c~=l-+-S-E.P_A_R_"'~_"_Rs,_*,_-1'--~:r~,: ~'-- I ',;!u-;.,-;.,- -11' --- - ----\- -- - ' - - F I

,, LJ Ji -Iii.'° iT i\. J I

I TlllOU.EY !IEAM\ I r

- TURBINE BUILDING SHEET-7 SURRY POWER STATION I

FIG. 10.1-10 OCT. 15 1 1970

.... H9-lf\J ..:Hl17171 I. 3 I

4 5 0 7 8 9 l I I PLENUM (~YP) I ! . !- i I ! .. *1 I ' I I I r- _L I~

- I I j__7 I

I ,{ I I /

I I I I !

I

- FW HTR TlJBE WITMCIRAWAL """i I 1.

I I I I I I I I I LL I i (T'"l ; I i I I I I i I I I I I

=

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- - Et.58~t..*

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-- _J cit _J L , L .: ---- F 1 I

r-,. 3

~ ,- I 3 EL42'0" I

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tL rf-d I

- f, I II I : D - /

p

,eoTIOH_s_u_P_POR_T---1,:111!... "o"" RtFE!tENCE ! T~e; ;,T-;;D;.:~~- ,- ; * ...,~ Jl<" ,::, I I '-l----- 1 I n /i~ m 11 I* I* _.,-b Cl "'11 I" I I I b.

l*CC-E*IC I I I -- - ______ ":, ~

~

EL 9"-r,," ! , WATER 'oPILLAGE TRENCH I"

. j y- Pl PE. TUNNEL

ri.-t LP HlA,-E.R.. ;1 1 111 t LP HEAU.R.

DRAIN PUMP I I I 1: DR.AIN PUMP

,,, ,,, ____ L!!:.1-_oJ i,.9c;'10 C.W OISCHARG"° J11 l*SD*P*lA I I l*SD*ll'*lS J,0.11448 J.0.115-48 1$---~~:-$-+$ ~~ 3~

I 9 t SV MANl~OLD, /,.------I---,, I

t. GEN LEADS\ A A ,'"r\

~

OIL ROOM ~ODS' EL "Cr-ti' TROLLEY BEAH.J: r-(1"Pl I I

--,

I I

COND TUBE----. WITHDRAWAL (TYP) 11

"""'I I I I I

I . I I I I ' I I I L ___ J

DWI FH-G,A.

I TURBINE BUILDING ( SHEET-8 SURRY POWER STATION _J
10.3.1-l 12-1-69 10.3 SYSTEM DESIGN AND OPERATION 10.3.1 MAIN STEAM SYSTEM The Main Steam System is. shown on Figure 10~3-l. The turbine generator heat balance for valves wide open and for the anticipated 2,546 MWt rating, is shown on Figure 10.1-2.

10. 3.1.1 Design Basis

. Each of the three main steam pipes is designed in accordance with the ASME Code

/ for Pressure*Piping, ANSI B31.l, for a flow of 3,722,641 lb per hr of steam at 1,085 psig, 600° F. The pipes are each 30 in. OD A-155, Class 1, Gr. CMS-75 c~rbon steel, 1 in. minimum wall, and join to form a connnon 36 in. OD header. Steam flows from this header through four 28 in. OD pipes_to the stop trip valves and the turbine. The main steam bypass system is sized to take the excess steam flow resulting from a 50 percent load rejection. This resul_ts in a 10 percent reactor step change in power and a, 40 percent steam dump, having a flow of approximately 4,454,000 lb per hr. This flow can be divided equally through the 8 bypass control valves, with each valve having a maximum capacity of 890,000 lb per hr at full load steam conditions.

.~. The 4,200 rpm turbine driven auxiliary feedwater pump turbine is design~d to operate at main steam pressures from 1,100 psig to. 600 psig to acconnnodate the

1.0.3.1-2. 12-1-6.9 possible main steam* conditions from full load to that operating condition w.here the Residual Heat Removal System can be placed in s.ervice. The 1,800 rpm t.urbine drives for the containment spray pumps also use main steam. Small bypasses around each inlet control valve to both the auxiliary feedwater pump t~bine and the containment spray pump turbines continually admit a small amount of steam to maintain the blading and nozzles in a warm condition for rapid rolloff. The main steam piping supports have been analyzed for turbine trip forces as well as the seismic criteria. Since the turbine trip result.s. in a, more severe shock to the piping system than the DBE, as set forth in Section 2.5, the turbine trip data were used in the design of the piping supports.. In addition, the system has been stress analyzed for the forces and moments which result from thermal growth. ~~ main steam piping within the containment annulus has been reviewed for possible pipe r~pture and sufficient supports and guides have been provided to prevent damage to the con tainmen.t liner and adjacent piping. 10.3.1.2 Description Steam is conducted from each of the three steam generators within the reactor containment through a swing disc type trip valve and an angle-type nonreturn valve into a common header. The steam passes from the header to the turbine stop-trip valves and governor valves. A steam flowmeter (venturi) inter-connected with its three element feedwate-r control system is located in the line from each steam generator.

10. 3.1-3 12-1-69 The motor operated nonreturn (stop-check) valves automatically prevent reverse flow of steam in the case of accidental pressure reduction in any steam generator or its piping and also provides a shutoff of steam from its respec-tive steam generator.

The swing disc type trip valves in series with the nonreturn valves contain swinging discs which are normally held up and out of the main steam flow path by air cylinder operators. If* a steam line rupture occurs downstream of the trip valves, an excess flow signal from the steam flowmeter releases the air pressure on the air cylinder and spring action causes these valves to trip closed, thus stopping the flow of steam through the steam lines. A total of five ASME Code safety valves are located on each main steam line outside the reactor containment and upstream of the nonreturn valves. Four 6 in. by 10 in. and one 4 in. by 6 in. valves are provided with a total relieving capacity of 3,725,575 lb per hr. Excess steam generated by the residual and sensible heat in the core and the Reactor Coolant Systemr"'tt; normally bypassed directly to the condensers by means of two 14 in. main steam bypass lines, which provide, a total bypass capacity of 40 percent of normal full load steam flow. Each bypass line contains a bank of four steam bypass control valves arranged in parallel. These valves are controlled by reactor coolant average temperature with provisions. to con-trol a portion of the valves with steam pressure. An uncontrolled unit cooldown caused by a single valve sticking open is minimized by the use of a group of valves installed in parallel.
10.3.1-4 12-1-69 All or several of the bypass valves open under the following conditions pt*o-vided a condenser vacuum permissive interlock is satisfied:

1. On a large step load decrease the steam bypass system creates an artificial load on the steam generators, thus enabling the Nuclear Steam Supply System to accept a 50 percent load rejection from the maximum capability power level without reactor trip. An error signal exceeding a set value of reactor coolant T minus T f avg re .

will fully open all valves in 5 sec. -T f is a function of load re and is set automatically. The temperature controlled valves close automatically as reactor coolant conditions approach their pro-grammed set point for the new load.

2. On a turbine trip with a reactor trip, the pressure in the steam generators rises. To prevent overpressure without main steam safety valve operation, the turbine steam bypass valves open and discharge to the condenser for several*l&inutes, to provide time for the Reactor Control System (Section 7.3.1) to reduce the thermal output of the reactor without exceeding acceptable core and coolant conditions.

3. After a normal orderly shutdown of the turbine generator leading to unit cooldown, the pressure controlled bypass valves are used to release steam generated from the residual and sensible heat for several hours. Unit cooldown is programmed to minimize thermal

10.3.1-5 9-15-71 transients and is based on residual and sensible heat release. It is effected by a gradual manual closing of the bypass valves until the cooldown process is transferred to the Residual Heat Removal System (Section 9.3).

4. During startup, hot standby service, or physics testing, the pres-sure controlled bypass valves are operated from the Main Control Room.

All bypass valves are prevented from opening on loss of condenser vacuum and excess steam pressure is relieved to the atmosphere through the main steam safety valves. Interlocks are provided to reduce the probability of spurious opening of the bypass valves *

During normal operation, steam from the main steam manifold is used for turbine shaft sealing and for auxiliary services.

An atmospheric steam power relief valve with an adjustable set point is pro-vided on each main steam safety valve header, upstream of the nonreturn valve outside the containment. The relieving pressure of these valves, normally 1,035 I psig, is individually controlled from the Main Control Room and each valve has a capacity of 373,000 lb per hr. In addition, a decay heat release control valve is provided which, after approximately 1/2 hr, can release the sensible and core residual heat to the

10. 3.1-6 12-1-69 atmosphere via the residual heat release header. This valve is pGsitioned from the Main Control Room.

The valve is mounted in the common decay heat release header and services al1 three steam generators through 3 in. connections on each main steam line. upstream of the nonreturn valve. In additionp this valve can be used to release the steam generated during reactor physics testing and operator license training, and while the unit is in the hot standby condition. A nonreturn valve is provided in each line connecting the main steam lines to the common residual heat release header to prevent reverse flow of steam. Steam is supplied to the turbine drives for the auxiliary steam generator feedwater pump and containment spray pumps from each steam line upstream of the main steam nonreturn valves. The steam lines to each turbine are con-tinuously under steam generator pressure up to the shutoff valves located at the turbine*drives. The motor operated steam supply valve for the auxiliary steam generator feedwater pump_ is operable from the Main* Control Room and the Auxiliary Shutdown Panel. Operation of this valve is also initiated auto-matically from a loss of power signal or on low-low level* signal in 2r!of the. 3 steam generators. Indication of all operating conditions is pro~ided in the Main Control Room to enable the operator to adjust feedwater flow with any of the 6 motor operated valves shown on Figure 10.3.5-2. Steam leaving the high pressure turbine passes through four moisture separator-reheater units in parallel to the inlets of the low pressure turbine cylinders

Each of the four steam lines between the reheater outlet and LP turbine inlet

10.3.1-7 12-1-69 is provided with a crossover stop valve and an intercept valve in series. These valves, operated by the turbine control system, function to control turbine over-speed. Six ASME Code safety valves are installed on each crossover line between the high pressure turbine exhaust and the moisture separator inlet to protect the separators and crossover system from overpressure. The valves are sized to.pass the flow resulting from closure of the crossover stop and intercept valves with the main steam inlet valves wide open. Although this event is uniikely, the valves discharging to atmosphere prevent equipment damage.

10. 3.1. 3 Performance Analysis If a main steam pipe rupture occurs '<section 14. 2. 8), a flow signal measured by the venturi flowmeter located in that main steam line causes the swing check trip valves in all three i;nain steam *lines to trip closed.

for the trip valves is approximately 5 sec. Minimum closing time If the rupture occurs downstream of the trip valves, valve closure stops the flowfof steam through the pipe rupture, thus checking the sudden and large release of energy in the form of main steam. This prevents rapid cooling of the Reactor*Coolant System and an ensuing reactivity insertion. Trip valve closure also ensures a supply of steam to the turbine drives for auxiliary steam generator feedwater pumps described in Section 10.3.5. If a steam line breaks between a trip valve and a steam generator, the affected steam generator continues to blowdown. The nonreturn valve in the ruptured ',. line prevents blowdown from the other steam generators. steam break accident and is discussed in Section 14.2.8. This is the worst
10.3.1-8 12-1-69 Tests and Inspection The turbine is overspeed checked during normal unit startup. During this pro-cedure, the turbine bypass system functions and atmospheric steam dump and residual heat release valves also operate. The auxiliary feedwater pump and containment spray pump turbines are tested each.month. During unit shutdown the tripping mechanisms for the swing check valves will be tested for proper operation. The nonreturn valves will also be tested to verify that they are in operable condition.
FIG. 10.3-1 OCT. 15, 1970 ON'E. CA,P ON EAC:.1-{ HE.ADER ,----- \4 -SHP-IS-GOI 10 0,E SH\PPEO LOOSE FOR FIE.LO WEU)IHG PRIOR TO CHtl.\lCAl CLE~ll*N~, STE.AM 1 GENE.~"'-,-OR.'5

----1------~

vo5-6oc('TVP)

-&\ , TA I *-30"-5HP-l-~OI . Mii02°'-~:~~

VOS.oGOC; :-- -fa .Fl- * ~

~ ~ .-vos-c;.oc (TYP) )~

I' ' 14T-'1

! TO ATMOS i (TYP)

LT~STGEN BLOWi TAMK FM-12 2A (TYP) N0.2 LP. TURBINE 1---- l"ORAl~-- FM-2.8A. 30..*SHP*2.*60\

! 10*0.P TO BE S~IPff:tl-

L005E F'lR Ftf.LD 1

WElDlt.:(t AFTEJC CHEMIC! I w

*~"

u, CLEA\.IIN(., 2"DRAIH FM-t8A(TVP.) O~DE:; INSIDE TURfl BLOS r--WALL CONDENSER CONDENSER 1-CN*St:* IA I-CN-SC-18 so"-;;_Hl::'*3-"ob &ov YJ~....-iM--""T'--:---------,,..aF"-R-0-M-C-OS-.T"'A-'lltMEt'IT-/. - - ~ _,/Msio9 llOLAl\OM 5YSTEt.\ .- -"' T V ~ M5109. 3"-S.I-IPO- 5-C,.O\ OUTSIDE. l)J'5ID£ TURB.BLD6.

,. l-!.HP-13-~lj 0-.....--+-'-------i----,..~-------J 18 12.°'*5HPD*\5*601 LI bRAIN FU*28A(TV!O)

VE>S*GOB Non:s: f'LOW' DIAGRAM SYMBOLS, STD HM-IOO TO H~!l-1081/\lCL 3~*51-1P*28*E,OI

--~--,------------------~~1------[>I $ SK-IC -753, REV 5-61. ' - - - - - ! - - - ___ ~f-sHP*_2_9_-s_o_,_ _ _ _ _ _ _ _ _ ~*-l-----=<r-t ALL PRES-SURE COIIINS TO SE f. 1 ALL TEMPE:.RATVRE (ONNS TO BE 12 REAMED HOLE, EXCEPT ,-\5 NOTEV 3"-SHP-30-601

lNO\CAl E5 FURNl~HED BY EQUIPMEl,,/,T MFR.

2ETC ~HOWN AT EQUIP CONN5 INDK.ATES MF'R!;) ti:>NN VCW*60A..l ~..l".. IMBEn - CUND ov;,::; N4-2250-11Jl,B'l-SODU.

- - F'RCM CONTAINMENT (TYP) -TUAB OWG 6"80.J387.

ISOLATION ~YGTEM f"GS*<oOB-

REACTOR CONTAINMENT PENETR.ATION'i:i 11448*FV*IA..

~ V ' VC5*(aOA. ! MSIIO 'll!: lit INDICATE:5 POW'-R RELIEF' VALVE MOOUlATII-IG TYP'E .6. l!IDlC:ATE~ cm.,PUTE.R PCIN.T MAIN STEAM SYSTEM . FKDM~E~ '-INITl"-"l\!ON 1:,'(!i z**SHPD-1'1*6C' A.UX ;: EEO PUMP DRIVE FOR AUX ~T \:JEN FEED PUMP (l*F'W-P-2)

UNIT I.

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.j 10.3.2 AUXILIARY STEAM SYSTEM 10.3.2-1 12-1-69 An Auxiliary Steam System is provided as shown in Fig. 10.3.2-1 and 10.3.2-2. All piping is designed in accordance with the ASME Code for Pressure Piping, ANSI B3Ll. 10.3.2.1 Design Basis The Auxiliary Steam System supplies 150.psig saturated steam throughout the station for the following auxiliary services: Condenser vacuum priming ejectors Condenser steam jet air ejectors Flash evaporators Steam ejectors for the chilled water units Air ejectors for the chilled water condensers Building heating and other building services Boron recovery system heat exchangers Waste disposal system heat exchanger Containment vacuum ejectors Containment spray pump turbine drive test steam
10.3.2-2 12-1-69 10.3.2.2 Description Normally, the auxiliary steam supply header receives its steam requirements from the, turbine crossover* lines. During. periods of low load operation when cross-over steam pressure drops below approximately 140 psig, steam is supplied from the main steam header through a pressure reducing valve. When both reactors are shut down, steam is supplied by the heating boilers. The Containment Vacuum System steam ejectors are used only during startup peiiods to initially evacuate the containment. During normal operat1.on, two mechanical vacuum pumps maintain the vacuum, as described in Section 5.3.4. The condenser vacuum priming ejectors are used during startup to draw the initial condenser vacuum. During normal operation, the steam jet air ejectors maintain condenser vacuum, as described in Section 10.3.6. Two heating boilers, each rated at 80,000 lb per hr of steam, are provided for preliminary and shutdown operation. Each boiler is the packaged water tube type and is equipped with motor driven fuel oil pumps, deaerator, and feedwater pumps. Number 2 fuel oil is supplied to the boilers from the main oil storage tank. Lines are provided !rom the auxiliary steam header to test the containment spray pump turbines. r --
10.3.2-3 12-1-69 10.3.2.3 Performance Analysis A loss of normal alternating current power will shut down the heating boilers. No services supplied by auxiliary steam are required to function as part of engineered safeguards during a loss of station power. Tests and Inspection Tests and inspection are performed on a periodic basis

FIG. 10. 3.2.-1 OCT.15,1970 V91-I/\J.:l-8i7i7 I I

~** TEMPOR:A'iI" CONI\J, CHEM- CLEM>ll"lG, FM-IIC,A TO C.ONTAIMME.NT -T() REACTOR CONTAH.JMENT ~PRA'< PUMP':. g AUX BUILDING (FM-\G,B) *1'2."

TUl<~INES (fM* 14,,._) 4",5A.",Q/ TO P~V STA ND. I {FB*"A) RA V~S-Goa

\/05*'2:oOC- IZ."*S,..*I* 301 u

2.~S.AD-301

-+"AJA.-1-1?.I s. ------~I ~'----~---.----~---~

0 r------'_t_*-s_A_-1_1-_,_0_1 3 vo:*toc.1 ,. z*. SAD *.!Cl VOS"5r*** *-----'~-f- A'SIOZ'-. 6-ACA-8*121, i ' Z r-----,-0--,---, ~----,-o-~- ;, .

J*SAD'".301 VGF-120.....-::

6H*ACA-1*1~1

'lt:l CO~CEN5Ei:: ;;; CO~N. 19 a, FM* 15A g z ~ ~ "'

I - ' - - - - - " - - - - - ~ :I

'ti,

. LTo C0ND£NSE.R L"

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El L

It CONN \9 ~o= ,*

*r:.:

(fl,H5A') VGF-12b(Cl\AIN OPEKA.TED' l*CN-EJ-2.A VOT-2.0G TO WASTE---! AIR EJECTORS

----* VOT-20G 'f TO WA':::.T'f.

ff VGF-1?..D(CHA.IH OPE.P.ATE.OJ I-C.N-EJ*2B* I IO'!.ACA-5*1'2.\ 12."*AJA*6-/2J. e: I EJECTORS i!I 1ci'-ACA*3*12.I VGf*l2D 4*SA.D-IS.I

': SAD*301 FROM LP TUR8\IJE @)DWG 71BJl68 7

GLAND STEAM NOTES:

C.ONDENSER ALL PRESSURE CONN-TO BE l* "

ALL TEMPERATURE CONN. TO BE I.

/ 7 ETC SHOWN AT EQUIP CONN. INDICATES MFR's / VGW*\SA *,, 7 CONN.NO. ON CO~DE.NSER DWG N4 *2.25D-RREBT

50\~J ft-.R)r

/ CONOENSER WATER SEALED VAC.UUM BREAKER VALVE CONDENSER '3<!-f-----~-,--,<

32,f----<a>-~ C0\..11\1, 19 FM*t*~ 5?ll.LOYE.~ FROM

---IJOT-20G *INDICATES FURNISHED SY EQUIP MFR. /J.. ':l~E I-K owe. C,4 5,50X.~l .

l*C-N:SC:*IA I* C.N .- Sf. *IB 32 ,___ _,,.,___ ___, ST. SEAL '5'l'STEM Lvo:, -Goe

yTOWASTE.

I @) CWG 118J1&8 CO~N NO 40 J O-VOT*20G

/ LS*' ---------!- [\ICT-'2.0E

_ _.,/ ~ L___~ E GLAND ST CONO a-WSC*l'll DRAIN PUMP 1-SD-P-3 R'.'c.TURN HOR. r'\1\-Z!A. AUXILIARY STEAM SYSTEM-SHEET I SURRY POWER STATION
FIG.10.3.2-2 D H OCT. 15, 1970 891-V'J.:J -8171711 aeD& LIN!. (l'1"1°l'fA) 1 **\105*C.OC. I PJ C,,A.S STRIPPER FEED /"ASI03 oTE.AM Hl<t>.Tl:Fc 2'B1l-75*1S2 O'V'ERFL.OW CA l"*SLPO*&B-151.-0-/ r-."1Bo2 1C "CID Y. / '.,,-~- ASIOO

/ AS IOO ! B.\TCH TANK

_(rM *IOSA) 4*-BR*?O*ISZ .;*-BR-92*!52 4*-sR-94-152

! 1-CH*TIC.*5 J" CL.ISi I "CLIS/

I f'Fl,\-2qA) S-IS'

}"*c.L 151 5*13 \ T*IS' T-13 2"-BR-~1-ISZ z'!BR*93*t5Z ' VC.S *<..OA 2"'*8R-95-152 VCilS*"O&

VC5-t.OA

.-t:.SA-2.Z-151 ----vos-c.oc l!,ORON &GRON .. _If.'i___

BR 111oB RECOV'cR'I' RECOVERY TANK T"MK 1-11::.AnR "EAHR

* '--;JC~*l89*151 (FM* Z2C)

2"-sA-aa-JOr s*-s,1-3<,-301 2: -SA-11H>OI

-2.'._ &LP0-&9-151 lz"-SLPD-"9*151 *-,a.pc,-a9-1s.

4""*SLPD*Sll*1S~ AUi !iT SUPPLY' H[lR VGW-30.A FUEl NORTH AUX.BLOC,. !!>LOG.. I *-SLPD-64-161 "l"-SA.\.1-30\ ii "IO PR.V STA. kO. 4 (A>*21D) ~

~ '4-'*SAM-301 TO PRV STA. NO. 3 t--01>-~--1 (Fe

21 D)

VGW*30A -*--------- v,w-3oA e'1CV*l09*1SI JOS*GOC 80CV*9*1~1 (FM-102A) 1---<ic1.. ,sr4a~ 3' 2-cv-~.i-1 CONTAINME!o!T \/'-CVUM EJE CTOR1i 2'"'-SLPD-55-153 3"*M-24*30J PRIMAAY W.-.rl:tll.TAMK H~TER I~ C.L 15"3 f\EFEREN:c DW6S~ T-10

&ORON uco~llY i,VS.TE ..

C.ONOEN,...TE

..-M*2.9A 1 8 l"M-17,' -.c W"&TE .16Jl'OaAL l'M*SOB S*IO AUX&Tt A I R - L FM-16A CONT 'l"C t LEAi< MO>llioll SVS Fll-101" D!.CON."T'~IN~ilON. SYS F"1-118A e.

CIIEl,H VOLUME CONTROL ;vs F'M*I06A COMPONENT COOLINS FM-22C. AUXILIARY e,oRON ot.VA.l"ORIIIT'DR NOTES: RE~O\LEll.5

LOCATED 1M cr;:ossoVER PIT ON NOA.TH SIDE OF" ROAD

*

LOCATED IN PUMP HOUSE STEAM WA.t>TE D15PC',,AL.

EVAl'OR.0...TCR. R~HDILER SYSTEM-SHEET 2 SURRY POWER STATION
10.3.3-1 12-1-69

10.3.3 TURBINE GENERATOR .

The turbine generator heat balance for valves wide open is shown in Fig. 10.1-1. The turbine is a conventional 1,800 rpm tandem compound unit consisting of,one double flow high pressure cylinder and two double flow Jow pressure cylinders. The turbine is expected to achieve a maximum capability of 855,408 kw gross with inlet steam conditions of 748 psia and 1/4 percent moisture exhausting to 0 1.5 in. Hg abs with a feedwater temperature of 441.1 F and 0.5 percent makeup. The turbine is provided with six stages of feedwater heating and four moist1i,1re separator reheaters located between the high pressure and low pressure*cylinders *

Each high pressure steam line to the high pressure cylinder contains a stop-trip valve and a governor control valve. Stop valves and intercept valves are provided at the discharge of the moisture separator reheaters to the low pres-sure turbine cylinders.

A gland steam sealing system is provided to prevent air* inleakage and steam outleakage along the turbine shaft. All necessary piping, controls, and a gland steam condenser. are provided. The hydrogen inner-cooled generator rating is 941,700 kva at 60 psig hydrogen gas pressure, 0.90 pf, 0.58 SCR, 3 phase, 60 CPS, 22 kv, and 1,800 rpm. Generator rating, temperature rise, and insulation class are in accordance.with the latest ANSI standards

10.3.3-2 12-1'.""69 Primary protection of the main generator is provided by differential current and field failure relays. Protective relays automatically trip the turbine stop valves and electrically isolate the generator.

A rotating rectifier (brushless) exciter with a response ratio of 0.5 is pro-vided. The exciter rating is 3,300 kw, 500 v d-c and 1,800 rpm. The exciter consists of an a-c alternator coupled directly to the generator rotor. The alternator field winding is stationary, and control of the exciter is applied to this wiµding. The alternator armature output is rectified by banks of diodes that rotate with the armature. This direct current output is carried through a hollow section of the shaft and is applied directly to the main generator field. The 22 kv generator terminals are connected to the main step-up transformer and the unit station service transformers by 22 kv aluminum conductors, each rated at 25,000 amp. Each aluminum conductor is enclosed in a forced air cooled, isolated phase bus duct. The turbine oil systems include a conventional design electrohydraulic con-trolled governing-trip system, and a low pressure bearing lubrication system as discussed in: Section 10.3.7. Oil is also used to seal the generator shaft seals and to prevent hydrogen leakage from the machine. Hydrogen side and air side a-c motor driven seai oil pumps are furnished to provide seal oil,for the prevention of hydrogen leakage from the generator. An a-c motor driven high pressure hydrogen seal oil back-up pump and a d-c motor driven, air side seal oil back-up pump are provided. A continuous bypass type oil purification system removes water and other contaminants from the oil.
I/ 10.3.5-1 12-1-69

;* 10.3.5 CONDENSATE AND FEEDWAT~R SYSTE~

The Condensate and Feedwater Systems are shown on Fi~ures 10.3.5-1 and 10.3.5-2 and the heat balance used for station design is shown on Figure 10.1-1. 10.3.5.l Design Basis The condenser hotwell is designed to operate at a normal level such that about 4 minutes of condensate flow (65,000 gpm) is available to supply the condensate pumps. A 300,000 gal condensate storage tank floats on the system. Each of the three vertical barrel type condensate pumps is rated at 9,000 gpm and 1,070 ft TDH. Minimum flow of approximately 2,000 gpm through the pumps and

gland steam condenser is maintained by an orifice measuring device downstream of the gland steam condenser. The orifice measuring device operates the recirculation valve as shown on Figure 10.3.5-1.

Two half-size steam generator feedwater pumps, each rated at 13,800 gpm and 1,700 ft TDH, are furnished to supply feedwater to the three steam generators. Each feedwater pump is equipped with two 3,000 hp electric motor drivers in tandem. Minimum flow through each pump is maintained by flow nozzles in the discharge lines. The recirculation valve opens when the flow drops to 2,800 gpm. Feedwater leaves the first point heaters at 440° F. A full-size turbine driven auxiliary feedwater pump rated at 700 gpm and 2,730 ft TDH, and two half-size motor driven auxiliary feedwater pumps rated
10.3.5-2 5-1-71 at 350 gpm at 2,730 ft TDH, receive suction from a separate 100,000 gal capacity condensate storage tank which is buried for tornado missile protection. The design is based on the following conditions:

1. Integrated residual heat release from a full power equilibrium core.

2. Feedwater inventory of the steam generators operating at normal minimum feedwater level.

3. Minimum allowable steam generator feedwater level permitted to prevent thermal shock or other damage.

4. The temperature of the feedwater which is supplied from the condensate storage tank. This temperature was assumed as 30°F when considering thermal shock, and 120°F when considering feedwater enthalpy.

The auxiliary steam generator feedwater pumps are located outside the contain-ment in a tornado missile protected enclosure near the main steam line and feedwater line containment penetrations. They take suction of 35°F to 120°F water from the 100,000 gal condensate storage tank through individual pipes which are contained in a concrete tunnel for missile protection.

10. 3. 5--J' 12-1-69

The pumps, drives, piping, and 100,000 gal condensate storage tank have all been designed as Class I components (Section 15.2.1).

10.3.5.2 Description* The .Condensate and Feedwater Systems are shown on Figure 10.3.5-1 and 10.3.5-2. Condensate is withdrawn from the condenser hotwells by two of the. *three half-size motor-driven condensate pumps. The pumps discharge into a common 24 in. header which carries the condensat~ through two parallel steam jet air ejector condensers and through one gland steam condenser. Downstream of the gland steam condenser; the.condensate passes through the flash evaporator condenser. The conunon header divides into two 18 in. lines which carry condensate through a pair of heater drain coolers and the tube side of two parallel trains of five low-pressure feedwater heaters to the suction of two half-size steam generator feedwater pumps. The steam generator feedwater pumps discharge through two parallel No. 1 feedwater heaters to an 18 in. discharge header for distribution to the steam g~nerators through three individual feedwater flow control valves, positioned by the three-element feedwater control system for each steam generator. A remotely operated small bypass valve is provided for manual control of feedwater flow to maintain steam generator levels during low power operation or hot shutdown. Drains from the moisture separators, reheaters, and the No. i and No. 2 feed-water heaters are collected in the high pressure heater drain tank and pumped

into the suct:f.on of the steam generator feedwater pumps -by one of two full-size high pressure feedwater heater drain pumps.

10.3.5-4 12-1-69 The principal controls of the Condensate and Feedwater Systems are located in the Main Control Room. The system is arranged for automatic or manual control. Impure condensate in the condenser hotwells is discharged from the cycle under administrative control through either a double valve connection to the circu-lating water discharge canal if activity levels permit, or through a normally blind flanged connection to a tank truck. A condensate storage tank is pro-vided for makeup. The amount of makeup is controlled by low hotwell level . . A recirculation control to the hotwell returns condensate at low generator loads and provides the minimum amount of water for the air ejector condensers and the gland steam condenser. A full-size auxiliary steam turbine driven feedwater pump supplies feedwater to the steam generators during a complete loss of station power. During

. periods of startup, and for core residual heat removal, two half-size auxiliary feedwater pumps, driven by electric motors connected to the station emergency power supply, are used. Feedwater flow to the individual steam generators from the auxiliary feedwater pumps is controlled from the Main Control Room by remotely operated control valves in the supply line to each steam generator.

Steam to drive the auxiliary feedwater pump tu;bine is supplied from the main steam lines inside tornado proof structures upstream of the nonreturn valves. The turbine driven auxiliary pump automatically starts on loss of power or on a low-low level signal in two of the three steam generators by the opening of the steam supply valve. This valve may also be opened remote manually from the Main Control Room. The motor driven pumps act as backup for the turbine
10.3.5-5 12-1-69

drive pump in the event that insufficient main steam is available. Feedwater is discharged to the steam generators through connections in each ~ain f~ed-

. water line inside the reactor containment. Check valves prevent loss of feedwater, should .a feedwater line rupture. Isolation is provided as described in Section 10.3.1.

Chemical feed equipment (Figure 10.3.5-3) is provided to add amine, phosphate, and hydrazine solutions to the steam generator feedwater inlet line. Hydrazine is added to the feedwater to control residual oxygen content, and cyclohexalyamine is added to maintain an elevated pH of approximately 8.5 to 9.0. These chemicals act as corrosion inhibitors to reduce pickup of metal by the feedwater. Phosphate is also added to'contro'i possible scaling on the secondary side with 1 the con-

- tent being adjusted separately for each steam generator. Solutions are mixed and stored in covered feed tanks and pumped into the main condensate and feedwater systems by two motor driven, positive* displacement pumps with manually adjustable strokes.

10.3.5.3 Design Evaluation Both half-size motor driven auxiliary feedwater pumps operate from the station auxiliary power bus, and the full-size turbine d~iven pump can be used for residual heat dissipation as long as adequate main steam is available. The steam supply lines to the turbine are continuously under main steam pressure to keep them warm, and to prevent the formation of water droplets on turbine startup.
10.3.5...;6 12-1-69 Steam traps are provided in lines to ensure that any condensate formed as a result of warming is removed. The turbine is a* single inlet single stage unit and any drops of water forming do not damage or impair its operation. When m~in steam pressure is no longer adequate to operate the turbine driven pump, the need for residual heat removal is reduced to a level wherein a half:...Size motor driven pump can be used if necessary. In the event that only one half-size pump is available to supply feedwatei:'* following a loss of offsite power, there is adequate capacity to cool down the reactor. The effect of this transien~ on the overall steam generator fatigue usage factor, as stipulated in Sec~ion III of the ASME Boiler and Pressure Vessel Code, is that the allowable fatigue usage factor of 1.0 is not to be exceeded. All auxiliary feedwater pumps receive suction from a separate buried 100,000 gal condensate storage tank which is maintained full at all ti*mes. Operation of the auxiliary feedwater pumps provide residual heat removal for up to 8 hours using this tank. An emergency source for necessary feedwater is the Fire Protection System main. The three auxiliary steam generator:feedwater pumps with redundant means of motive power and associated piping are installed in a tornado protected area adjacent to the containment so that their use can be relied upon during any loss of station power accident. 10.3.5.4 Tests and Inspection The auxiliary steam generator feedwater pumps and'.drives are tested monthly by admitting steam to the turbine drive or energizing the motor drivers. During these tests, the pumps circulate water from the condensate storage tank through a recirculation line and back to the storage tank.

10. 3. 4-8 12-1-69

TABLE 10.3.4-1 AUTOMATIC OPERATION OF CONDENSER AND SERVICE WATER VALVES Initial Valve Action Accident Service Water Valves Main Condenser Valves Design Basis Loss-of-coolant (a) Open recirculation (a) Close all valves, either unit and loss spray heat ex- both units of station power changers to the affected unit (b) Close all others Loss-of-coolant (a) Open recirculation (a) All remain open, either unit with or spray heat ex- both units without a loss of* changers to the power to affected unit affected uni ts

.. Total loss-of-station power (b) (a) All others remain open Leave recirculation spray heat ex-changers closed, both units (a) Throttle all valves, both unj_ts (b) All others remain open

"FIG.10.3.4-1' OCT.151970 IZ- V'l..:l-917"17 I I 0:wa10-2.-1s1 i::t,.-1-1'22A i=LOWDOWN TMJI< DRAlf..l "--1H1S E:QLIIPME.NT TO BE. LOCATED ,t,T THE 'SHL F'II OISCHARG!: TU!,JNEL rrT *REc1Rcl.ll.A.T\ON 5PR~v l--l'i.ti..T C.'(CHt..NGE.KS RS-cit,,,

V _t; t-R5-E*l'B \*RS-E*IC. l*R5-E-IO _~fwr *

-20<,

(TY9) cuo,~r ,* FROM AIR CONO Q H U,JtT(r'9*41A)

'\. ~" TO SLUICE TREN0-1 0

(FP-71A) J ~" 4"WCPO- ?>0:,-!0\ (FM-\1A) I t MOV

~ ~

fRONICONO'ENSAiE 5'1'.S F'ROM (.ONO PUMP 4, REC (FB*.!!,A) To ~~~ 3 .. WI-IC>-IZ\ FROM Dl:.MINERALIZED

- _6~ WATE.RAR'EA SUMP PIJMP VVF-SA "*WC.W-1*\~b ,(FM*rTAY (TYPJ ~

a Q Q

~ 1~

ru ~

.;, ~ ~ ~ *FE SW!OCoC.\

C\~LA'TING WATER PUMPS

~ CAPACITY- 2101000 GPM SWIOSB l!EAD *?8'TbM \_

f(T'fP) NOT'E (f SIZE) COND CONN NO'S SWIOSli., 47,48,.51,E:; 52. TO BE 1t 150 LB FLG. NO 46 TO BE f ISOLB 1-CW-P..IC FLG. Q t0' UNIT I ti COMPONENT COOLING H'2..0

~ l*C'lll*P*\D HEAT* £)(CHANGERS ;~

24°'Mf1 1-2*'ws-13-10 4z'*WS-i<-IO r-----

UNIT I 1 L_ _ _ _ _ _ SEARfNG COOL\t,IG H ..O Ha,111.T EXCH .. '-lei&:JilS M V " C1.J.ILLE.D WA"TER SWIOIB 48 WS* I* 10 "\JNIT C.O'NDEN.~'f.R._ 1-CD*SC:.-t VC.F*l'i!:O 4e"ws-c.-10 NOTEC:I: ...

&, DENOTE~ COMPUTER POIHT&

P.IPK 51ZE .FOR *30" TO 96- PIPINC:, IS l>t'illOE.D!AMtT~R. PIPING TO BE IN ACCORDANCE WITH6PECIFICATION l'OA . WELDED CIRCULATING 'M'.\TER PIPING~CEXCE.PT /1,,S'. NOTED) 2, rr,.sH=iwu* AT EQUIP CONNS INDICATES MFrfs_ CONN NllMBE R - C.OND DWG N4 *225 0-RREBJ.S'OIXI

- REACTOR CONTAINMENT PENE1"1t,\,TIONS \l+lBfV~IA.. -

VOT-Z.OG ALL PRESSURE"CDNNS:TO BE f, EXCEPT AS NOTED. (TYP) ALL TEMPERATURE CQNNS TO.BE 1: El<CEPT AS NOTE"C.

.._-..J:L ,.

FUFH,ll~HED 'e.Y F.QlJlPME.NT MFR CWIOSD 3.°-W 5*!1?.*l3' TO INITAl(E tlllAL 4"-wsw*7-t36 vv<;-~2.P~ z.* Efll\ER6ENC..Y SERVlC.E.

SE.RV!CE \'J'A"TEt: PUMPS WAT?.R PUMPS Et-lC:IINE D~IVEN l*S'H-P- 4A.!48 CAP iSOOO Grt'1-E!4S" FT TD,H CAP 2800 GPM@15FT TOH (t'Ull 51?.E.) LA "flllP) I ROADWAY DE51GNE.O l'O Wln-15TAW I ~ LA~LOW) LOAD OF 2.00TONTR.ANSFJRMER. I I I

~~013, LA ~\G.W) ON LOW 2,~0 HlUCK CWIOIA Cl RCULATING LI I

j I

! I I

I , c LA WIOI \l',lTA\"<.'c. CANAL WATER SYSTEM c'_' ______l_ ___ J ______

- ___ _j_ .-:-:i+_,1' C_ ______________J L ---------,,.'-c------------~-"'--------------'FCLAP-'----'VJ.J:=VE=.S_ _.,__~

SURRY POWER STATION
10.3.3-3 12-1-69

An Automatic Turbine Startup function is provided to permit starting and shutting down the turbine with the aid of the Computer System. This improves the speed, consistency and reliability of the operation.

The program supervises steam temperature and pressure, steam temperature rate of change, turbine metal temperature rate of change, condenser vacuum, gland seal variables, differential expansion eccentricity, and vibration. Based on the information obtained from these variables, the program permits or rejects rolling off the turning gear, determines*permissible acceleration rates, and sets accel-eration rate through the EH governor. The program also computes, counts down, and prints out the heat soak time

10.3.4-1 12-1-6'.9 10.3.4 CIRCULATING ~ATER SYSTEM

The Circulating Water System, Figure 10.3.4-1, provides cooling water for the main surface condensers and the Service Water Systems of both units.

10.3.4.1 Design Basis To prevent the dir.ect recirculation of the heated circulating water dis charge, th~ system is designed to. take water from the James River on. the east side and to discharge to the James on the west end of the site. The shore line distance between the intake and discharge points is about 5. 7 miles, and the overland distance across the peninsula is about 1.9 miles. Each unit requires 840,000 gpm of river water to supply condensing and service water needs. To provide operational flexibility, system reliability, ,and station

,1 economy, the water requirement for'each unit is supplied by four 210,000 gpm pumps. These pumps qischarge to the common high level intake canal which con-veys the circulating water to the station area. Coarse trash is removed from the circulating water by trash racks at the river intake structure, and finer trash is removed at the station end of the high level intake canal by traveling water screens in the high level intake structure at each unit. The circulating water flows by gravity from the high level intake canal through four buried parallel lines to each condenser and then through four separate lines to a concrete tunnel for each unit. The tunnels terminate at seal pits located at the edge of the circulating water discharge canal, which is common to all units.

10. 3. 4-2

~ The discharge canal conveys the flow to the James River.

12-1-69 The discharge channel within the river is provided with rock groins along each side to control sedi-mentation and to maintain exit velocities of the circulating water ~o achieve desired dilution effects of the heated effluent. A circulating water discharge canal control structure extending partially across the opening between groins is provided to ensure proper velocity conditions when only one unit is operating. Sane components of the Circulating -Water System are used for handling service water, and are therefore designed as Class 1 structures and components. These components are:

- *The circulating water.intake structure at the river High level intake canal High level intake structures Buried circulating water piping and valves between the high level intake canal and the circulating water discharge tunnel Circulating water discharge tunnel Seal pits*

10.3.4.2 Description

\*-,

The circulating water is withdrawn from the James R:I:\ r through~ channel dredged in the riverbed between the main river channel and the shore, a distance of approximately 5,000 ft. The channel invert is 150 ft wide at El. -13.3. The channel is also used for shipping materials and equipment to the permanent dock on the east side of the site.
10.3.4-3 12-1-69 The circulating water intake structure is located at the shore end of the river intake channel and is an eight-bay reinforced concrete structure. The exposed deck of the structure is at El. +12; however, the pillbox enclosure for *the service water pumps is protected from flooding to El. +21 and from wave run-up to El. +33.5. The invert of the intake structure is at El. -25.25. Each bay houses one of the eight circulating water pumps for the two units. These pumps are rated at 210,000 gpm at 28 ft total dynamic head when running at 220 rpm. Each pump is driven by a vertical, solid shaft, 2,000 hp, induction motor. The pumps are of the non-pullout type and are serviced using mobile hoisting equipment. Before entering the pumps, river water passes through a trash rack at the mouth of each bay. This trash rack is _serviced by a movable trash rack rake which-dis-charges collected trash to a trough which then sluices the trash to the river. Each circulating water pump discharge line is a 96 in. diam steel pipe which conveys the water over the embankment of and into the high level intake canal. At the crest of the canal embankment, the crown of the pipe is provided with a pair of vacuum breakers and a tap for the vacuum priming system.* The vacuum breakers open when the circulating water pump is de-energized. These vacuum breakers and the flap gates which are at the discharge end of the 96 in. pipes prevent loss of water from the high level intake canal by siphoning through idle pumps. The vacuum priming system prevents air accumulations in the pump discharge line while the pump is operating. This system is isolated when the circulating water pumps are de-energized.
10.3.4-4 12-1-69 The high level intake canal is about 1.7 miles long and is designed to convey the circulating water flow of a four unit station. The canal is paved with 4~ in. of reinforced concrete, to allow velocities which would otherwise erode the earthen materials through which the canal is constructed. Since these earthen materials have low permeabilities, significant loss of water through the canal lining and into the substrata is not considered probable. 'l.;he bottom

)

( width along most of the length of the canal is 32 ft, and the canal h~s -/'::;~e slopes of 1~ horizontal to 1 vertical. The invert elevation variel- from

~-,

El. +S at the station end of the canal to El~ +6.8 at the rive:: end of the cana{~*- The benn along each bank of the canal is at El. +36.0. / I The water levels in the canal are controlled by the piping system friction losses within the power station and the prevailing river level. The normal water elevation at the power station end of the canal will vary between El. 121 and El. +23>> depending upon the tide. A minimum freeboard of 10 ft is maintained between the canal water surface and the berm at El. +36 during hurricane flooding of the river. This freeboard is adequate to contain surges in the canal which

*j..,.

could occur with a loss of station power when the river is in flood and is _main-/ tained by progress:J.vely reducing the number of pumps in operation by manual con-trol as the James River rises above El. +S .O. A reinforced concrete, high level intake structure is provided in the high level

'~

intake canal at each power* station unit. Each structure contains £ou{1 bays, and

,; )'

each bay contains a trash rack, a traveling screen, and an inlet to a'196 in. diam condenser intake line. Steel skin plates can be placed on the stop log I supports to permit unwatering of individual bays of the structure. The trash
10 .*3,. 4*-s 12-1-69 rack bars are 1/2 in. wide and are spaced 4 in. center to center. water screens have 14 gage wire with 3/8 in. clear openings. The traveling Screen wash water is* supp lied by two pumps , each _rated at 850 gpm at a 220 ft TDH. The four 96:in. diam lines connecting the condenser and the high level intake structure are reinforced concrete in the station yard and welded steel encased 1

,1 I .

in concrete under the station. Service water system taps are made in the steel

-*'-_;.,',...'{

port:(.m of \these lines.

/' \1, f ' ** ~-~< < J \,.- *L Electric motor ope:)~~te4_ butterfly valves are provi,ded at the condenser inlets * . *-'{:c .-'.

An Amertap condenser tuh~ cleaning strainer is installed in each of the four condenser discharge lines between the condenser discharge nozzle and the motor operated condenser dis charge butterfly valves. at the reinforced concrete discharge tunnel, which then carries the water to the common circulating w~ter discharge canal. These dis charge lines terrnina te This tunnel is 12 ft-6 in. by

-'12 ft-6 in. in cross seetion. */ [ . The circulating water system total energy gradient in *1:h~ discharge system is maintained at proper elevation to ensure a full con-denser discharge water box by a seal weir at the termination of the discharge tunnel.

The ~ischarge canal is excavated in earth and is designed to carry the flow of the qqo units with a velocity of about 2.2 fps at mean low water. The invert .J of tl)e canal is at El. -17. 5 and the sides slope at 2 horizontal to 1 vertical; I _this slope is stable under the Design Basis Earthquake Condition. The bottom width of the canal varies between 20 ft and 65 ft.

10. 3.4-6 12-1""."69 The discharge *canal extends about 1,200 ft into the James River. This extension has rock-filled groins along each side to minimize siltation and to provide the means to maintain a 6 fps terminal velocity of the discharge water. The opening between the groins is sized to ensure p~oper mixing of the discharge water with the James River. A timber pile trestle having five 10 ft wide bays in which timber gates.may be placed extends about half-way across the opening in the groin. The timber gates may be installed in this*structure using mobile hoisting equipment to reduce the net area of the opening between groins and thereby maintain the 6 fps terminal flow velocity when a unit is taken out of service *

10.3.4.3 Performance Analysis All four circulating water pumps for each unit should normally be in service.

If a circulating water pump is out of service, unit operation can be continued, but the station operator ~ust maintain a satisfactory water level in the high level intake canal by throttling the condenser outlet valves. The condenser inlet and outlet valves are normally controlled from the Main Control Room. When a consequence-limiting safeguard initiation occurs and there is a loss of station power, both inlet and outlet valves are closed, so that if one fails to close, the other will close *. The*valves are closed to conserve water in the high level canal for cooling the recirculation,spray heat exchangers.

When a loss of power occurswithout a consequence-limiting safeguard initiation signal, the condenser outlet valves are throttled to conserve water in the intake canal for the bearing cooling heat exchangers and component heat exchangers., and

10. 3.4-7 12-1-69 to provide-a minimum flow required by the main steam bypass system. If the

, water level in the high* level intake canal drops to El. +18, both the con-denser inlet and outlet valves are closed t.o conserve water in the high level intake canal for subsequent use. As mentioned in Section 10.3.4.1, certain components of the Circulating Water System are designed as Class I structures to preclude system failure during an earthquake,. and. are also designed to with-stand a tornado in order to ensure a supply of service.water in the event of an accident. The traveling water screens have been sized to prevent' trash from plugging critical heat exchangers in the Service Water System. Automatic operation of the condenser inlet and outlet valves and the valves in the service water system under various accident conditions are listed in Table 10. 3.4-1. 10.3.4.4 Tests and I'nspections Automatic operat~on of the condenser and service water motor operated valves, as described above in Table 10.3.4-1, are checked during initial operation and at frequent intervals thereafter.
FIG.10.3.5-1 OCT.15, 1970 I A C D E F 11 G H K

\tLHI\J~-817171 I A.1C ___ _

FEE.C WATE.R HE.AT~?.~ 5,S\1)0-& c~

~-s*o'--c_c * :::s --~_f. __ _ --i\ *-=:,,, 100-5 1-Ct-J*SC*\A

___c_c__ i CR

$Sl00_.... SSIOO -1

____!_'!..£_ 5SI00-7 SIJIBA. __l._!_~_ _/ S5 !DO*&

2 V Pl

/ TES.T -- vos-ai:,c w,

CN\i4A

~=* VGW*30A*G clB"-WCPIJ-26-301 D b VG.W-30A-G ~

3

* <iL"ND 5TE"M ;

COt-.lOE~SER VG.W-30A-G 14':.WCIPD-"a'l-301 PA V0S-60G c~,

~-*1*** ~ C.W\07 (CTYP) 5-ru. POH,11 HEAiER V6W-3Q#II.-G 1-FW*E-SA r---<><>---; l*CN*EL-1A t--<'Q----r-1 j i \8~~CPD*8*!01 ..D ..,*..,,-i I 0 1e*:wcPo-a2-ao1 TO W~lE STEAM C:i!:Nl!.RAtOR VOS*bO(; ('NP) 'F'EEC P\JMPS "TH POINT \-\E.A.i'E.1\

13,BOOGPM EACH@ 1-l=W*E-6B

F,oM MTR ORA.IN P\JMP(FM*\9-'J J,700 FT. TDI-\ (HALF SJ'ZE) 3\lfCPD*l!I co ...DENSER CON CENSER

.r,: I-CN*SC-1"- 1-CN.- SC:*\B OVEI\F\..DW -VG.F-ltl> C'..\5'!. CLIS\ .VGoF*liD 300.000 GAL v.o~-aoc.

COND'ENSAIE.

~TO'RA.Q~ I/lo.HK NOTI:.S:

I-CN-n:-2 HOTWELL EQUALIZER PIPES* I /;. IIIOICA'Tc'5 Ctl1'1'1lml Fl'.ll>ITS

TS~P C.P (CROSS*TIE ALL PRESSURE CONN'S TO BE i,.

TD UNIT*! CNDS MA(E*UP) ALL TEMPERATUR~ CONN'S TO BE t.

\1546-FM-\7,6,. - -~*--- 3 ETC SHO'NN AT EQUIP CONN'S INDICATES* Mf"R'S CONN NUMBER ON CONDENSER DWG N4-22'SD-A.Rt~T*501XI, 1s*:..wcPo-1-=oo1

INDICATES ~URNl~HED av EQUIPMENT MFR.

*7 'J~S-~oB -°"° '-"'-------, TO E)(HAUST HOOCl SPRAY \ , CONN 73 ON @)OWG 718JIG9 VOS-Gioe 1 --~ TC-CNI09AJ CONN43 f

TC -, @ OWG.

VOW*30A lf'2j.i;C1<Ui'.i4'.;f'\L-,;;~~<~N~lo~*~g ~-FROM TC-CNI09B 118.Jl"9 FOR ORA\t-Jlr-..)6

C.DNDC.\v'=i~R If' coNTA.M\t,..IA"TEO

\IOS*<.OC.('N'P) /

(PUMPED OUT 6'l' CCND. ?UM? iOTAJ.lk' i'RUC.K') - 3"CONt.JfCR"'OblHl'.:i ' :TO ~ISC.H 'Tlltu.ltL C."Et.llC.\LS (f'l'll*Z.lA) Pl

\J(;l'* 120 (TY?) \ TEST Lvos-6oc C.ONDENSA.TE ~~~PS, 9000 GPM EACH CONDENSATE SYSTEM @ 1070 FT TD7(HALF SIZE)

SURRY POWER STATION
50 - ;)-'6{)/ __,,,,,., FIG. 10.3.5- 2 /;di OCT. 15, 1970

'v'9t-V\J.::l-9171711 STEAM GENERATORS PT F EEO WATER HEATERS F'/-1100~

l*W:W*E-IA H='W-E-10

~RTO !4.*WFP0*17*1,,0I Fwi'i'iA \ - - - l'iOLAT ION VALW'I LEAK TE!.T V05*(,0C(1YP) ~

1* I; _.r:~~r~A I Fl

- ---J-l'YOS*fiOC "*WFPD*IS-601 CL601 CL901 r-(TYP)

VGS-60ll , \ ~___) l° SAMPLE CONN TO MOV Si\.MPLE SINK 'F'M*3l8 f-TEFUDF" VCW-60A- 14°-WFPD-a?- 901 1487 \ FWl548 PT~ Ft.1-IZ.IAt--,'---=.,,.,==,-, 1 r'""--~FW.;;l,;,5~3 FWISl ,a:wrPD-5

901 V0S*60C- l"VOS*ISOF \ '

~

(TYP)

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FIG.10.3.5-3 OCT.15,1970 H v'IGI- V\l..:l-8vvll LA 1-----------:,,' It. Cf$*301

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't- *-.-

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I I-WT-Tl<-3 I* RC* E*IB C y ., HYDRAZINE' TANK

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":r~ l"CL.301 .3 (FM*16A) PH05PHATE 'TANK * (CL '8.01

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COUNTERS If EQUIPMENT BY VENDnP.

l*RO*E*IC 0 V01,1*500H 10 BE f WHllt.lE.'(

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sov* WTIOOD CHEMICAL FEED SYSTEM SURRY POWER STATION
10.3.6-1 12-1-69 10.3.6 CONDENSER Two single-pass, divided water. box condensers are provided. Each condenses steam from one of the two low pressure turbine exhausts, and steam from the turbine steam bypass valves, as described in Section 10.3.l. 10.3.6.1 Design Basis The design parameters for each condenser are as follows: _Steam condensed, lb per hr 6,195,000 -~ Circulating water, gpm Surface, sq ft Number of tubes 773,000 580,000 55,624 Tube material 90-10 Cu Ni Tube OD, in. 1 Effective length, ft-in. 39-10 Back pressure, In. Hg 3.5 Temperature, °F 120 10.3.6.2 Description The condensers are of conventional design, manufactured by Ingersoll-Rand Company and have a Neoprene lined rubber belt type expansion joint in the neck. .. They also have steam and condensate crossover ducts to equalize pressure, impingement baffles to protect the tubes, and provisions in the hotwells for detection of circulating water inleakage.
10.3.6-2 12-1-69 The condenser hotwell is of the*deaerating type capable of reducing the oxygen content to less than *.005 cc per liter. The deaerating capability is necessary as there is no deaerating feedwater heater in the feedwater cycle. One No. 5 feedwater heater and one No. 6 feedwater heater is* located in each condenser neck. Two. twin-element', two-stage, steam jet air ejector units, each complete with tubed inter-and-after condensers, are provided for removing noncondensabie gases from the condenser shells. For normal air removal, one element of each ejector unit is operated per condenser shell. The ejectors function by using auxiliary steam and discharge to the ventilation vent. A radiation monitor is installed in the common discharge line (pf the two air ~jectors) to the.ventilation vent and is described in Section 10.3.8. For initial condenser shell side air removal, a noncondensing ~riming ejector is provided for each shell. These ejectors function by using:steam from the station heating boilers (Section 10.3.2). The condenser is equipped with an Amertap tube cleaning system complete with controls, piping, recycle pumps, collector, and sponge rubber balls.
10 . 3.6-3 12-1-69 10.3.6.3 Design Evaluation Loss of normal a-c power causes the four 96 in. condenser outlet valves to close to approximately 10 percent of their normal opening. This closure permits the minimum flow of circulating water to continue through the condenser for the main steam bypass system and conserves water in the intake canal for the recircu-lation spray coolers. 10.3.6.4 Tests and Inspection During unit startup and subsequent shutdowns, power failures are simulated to check the operation of the condenser discharge valves and the ability of* .the

I '

condenser to handle the maximum bypass steam flow at the corresponding circu-lating water flow.

10. 3. 7-1 12-1-69

10. 3. 7 LUBRICATING OIL SYSTEM A pressure lubricating oil system is provided to perform the following functions:

Store lubricating oil. Supply ofl to and receive oil from the turbine generator oil reservoir. Purify a side stream of oil from the turbine generator oil reservoir on a continuous bypass basis. Clean and reclaim used oil from the storage tanks, pumping it from the "used oil" storage tank via the purifier to the "clean oil" storage tank. 10.3.7.l Design Basis The lubricating oil system consists of a 21,000 gal reservoir, two 22,000 gal horizontal all welded steel storage tanks, an oil purifier, and two identical motor driven transfer pumps. The two gear type positive displacement transfer pumps are each capable of 2-speed operation at 108 and 48 gpm to accomplish* the various batch cleaning, _transfer, and circulating operations. The three compartment Bowser oil purifier is rated at *3,000 gal per hr. 10.3.7.2 Description A.turbine shaft driven oil pump normally supplies all lubricating oil require-ments to the turbine generator unit. An a-c motor driven turning gear oil pump*
FIG.10.3.8-1" C D L OCT.15,1970 A B t INSIOE REACTOR CONTAINM':_~_2°__ ~0rl 0UT5ID£ REA.CTO~ CONTAINMENT N51D~ REACTOR CONTAINMf..NT NO 2. STEAM STEAM GENERATORS GENERATORS

/ (

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(FM-32.Bi lFM-32A _ _ _ (TYP) 1 I 0

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10.3.9-1 12-1-69 10.3.9 BEARING COOLING WATER SYSTEM The Bearing Cooling Water System supplies cooling water to the Steam and Power Conversion System equipment and is a closed cycle using pumped condensate as cooling water. The heat removed by the cooling water is transferred to service water in the bearing cooling heat exchangers, as described in Section 9.9. The Bearing Cooling Water System is shown schematically in Fig. 10.3.9-1. 10.3.9.1 Design Basis The turbine plant equipment is designed for full load operation with cooling 0 water supplied at a maximum temperature of 105 F. The bearing cooling water heat exchangers consist of three half-size units capable of maintaining the cooling water supply temperature below 105° Fat all river water temperatures. The principal equipment served by the bearing cooling water is as follows: Equipment Design Flow, gpm Generator Hydrogen Coolers 6,840 Hydrogen Seal Oil Coolers 360 Turbine Oil Coolers 3,380 Exciter Cooler 400 Isolated Phase Bus Duct Air Coolers 150 Instrument and Service Air Compressors 30 Sample Coolers 50 Condensate, Feed, and Heater Drain Pumps 40 Flash Evaporator (During unit shutdown) 10,000
10.3.9-2 12-1-69 Two full-size 13,000 gpm motor driven pumps circulate the cooling water through the above equipment and the bearing cooling heat exchangers. 10.3.9.2 Description The cooling water flowing through the major equipment coolers, such as the hydrogen and oil coolers, is controlled automatically to maintain constant temperature of the cooled fluid. A head tank is provided to maintain a positive pressure at all points on the system. Makeup to this tank* is normally from the gland steam condensate drain pump; however, when this system is not in operation, makeup is provided from the condensate system (Section 10.3.5). * **** _\...-

/

10.3.9.3 Performance Analysis The instrument air compressor is the only piece of equipment requiring bearing cooling water during a loss-of-coolant accident or loss-of-station power. Under these conditions, jacket cooling water for this compressor is supplied by well water from the Fire Protection System.

..-~ ..

FIG. 10.3.9-1 a K OCT_ 15,19('0

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HP FLUID R'ESER'v'OI R BEARING COOLING OIL COOLERS WATER SYSTEM SURRY POWER STATION
10.3.7-2 12-1-69 is installed for supplying lubricating* oil during startup, shutdown, and standby conditions. An emergency d-c motor driven oil pump, operated from the station battery, is also available to assure lubricating oil to the bearings~ Cooling water from the Bearing Cooling Water System (Section 10.3.9) is used for the turbine lube oil coolers, which are immersed in the main oil reservoir. The two 22,000 gal storage tanks are normally designated "clean" and "used," but are interchangeable and are located inside a fireproof room equipped with water sprays and vent fans. The transfer pumps and piping are arranged so that oil can be processed from the oil reservoir or either of the two storage tanks. The processed oil can be returned to either of the other two. Vapor extractors purge oil fumes from the oil purifier and reservoir and exhaust to the atmosphere outside of the turbine building. lul piping and valves in the system are of welded steel, and high pressure bearing oil piping is enclosed in a guard pipe. 10.3.7.3 Design Evaluation Only the d-c bearing oil pump is required to function during a loss-of-coolant accident or loss-of-station power. The station battery provides* an uninterrupted source of power to this pump. No other part of the Lubricating Oil System is required to operate. 10.3.7.4 Tests and Inspection The d-c bearing oil pump is tested monthly.
10.3.8-1 2-13-70 10.3.8 SECONDARY VENT AND DRAIN SYSTEM -- Because the Steam and Power Conversion System is normally nonradioactive, vents and drains are arranged in much the same manner as those in a fossil-fueled power station. However, because air ejector vents and steam generator blowdown can possibly become contaminated and because they discharge to the env.ironment, they are monitored and discharged under controlled conditions as explained below. The air ejector vent and steam generator blowdown subsystem are shown in Figures .10.3.2-1 and 10.3.8-1, respectively. 10.3.8.1 Design Basis I . I Each of the condenser steam jet air ejectors (two per shell) is designed to remove 12.5 cfm of free air. Each ejector normally uses about 800 lb per hr of steam at 150-200 psig from the auxiliary steam header, while using 900 gpm of condensate for cooling. Separate hogging or vacuum priming jets are used to reduce condenser vacuum to 1-3 in. Hg abs during startup. The three steam generators associated with one unit are expected to collectively blow down 10,500 lb per hr of steam to that unit '.s blowdown tank under normal operating conditions. This tank is vented directly to the atmosphere.

-~

10.3.8-2 2-13-70 10.3.8.2 Description Vent gases removed from the condensers by the air ejectors are normally dis-charged through a radiation monitor (Section 11.3) to the atmosphere through the process vent. If a steam generator tube ruptures, with subsequent con-tamination of the steam, the radioactive noncondensable gases would be detected by the radiation monitor located in the air ejector effluent line. The related accident analysis is covered in Section 14.3.1. When the radioactivity level reaches the alarm set point of the monitor, trip valves in .the air ejector effluent line will automatically actuate to divert the effluent flow to the containment and shut off the vent to atmosphere. Other vents from the turbine generator which handle carbon dioxide, hydrogen, oil vapor, and other non-radioactive gases are discharged directly to the atmosphere outside the turbine building. Generally, secondary plant piping drains to the condenser. Each steam generator is provided with blowdown connections for shell solids concentration control. The rate of blowdown from each steam generator is con-trolled by a manually operated needle-type flow control valve. A blowdown slip stream is taken from a point ahead of each flow control valve to produce a composite sample for radiation monitoring. Blowdown from a unit's three steam generators passes to and flashes in the steam generator blowdown tank associated with that particular unit. The flashed vapor is discharged to the atmosphere through the process vent while the condensate is normally drained by gravity to the circulating water discharge tunnel and, when contaminated, to the vent and drain system (Section 9.7).
10.3.8-3 2-13-70 10.3.8.3 Performance Analysis 3 If the radiation monitor detects contamination exceeding 3.5 x 10- uc cc' in the blowdown sample, an alarm is initiated in the Main Control Room. At this signal, the operator shuts off all blowdown in the affected unit and drains the associated steam generator blowdown tank to the Vent and Drain System (Section 9.7). Individual steam generator blowdown samples are monitored separately to determine which steam generator is leaking. Loss of power or air causes both diversion valves in the air ejector line to fail closed, thus preventing possible radioactive contaminants in the condenser steam space from reaching the atmosphere. In addition, the air operated shutoff valves in the steam supply lines to the air ejectors will also go closed on a loss of power or air. Radiation monitoring, steam generator blowdown, and alarm initiation are unaffected.by loss of power, but a signal from the containment isolation system (Section 5.2), causes the trip valves on each side of the containment wall to close. 10.3.8.4 Inspection and Testing The Vent and Drain Systems are in continual use and require no special testing and inspection. *However, the trip valves installed in these systems, which are part of the containment isolation system, are tested in accordance with Section 5.2. **--**}}

ML19094A132: Difference between revisions

Latest revision as of 21:54, 16 March 2020

Contents

Text

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

2.2 DESCRIPTION

4.3 DESCRIPTION

5.2 DESCRIPTION

6.2 DESCRIPTION

8.2 DESCRIPTION

9.2 DESCRIPTION

10.2 DESCRIPTION

SUMMARY

2.2 DESCRIPTION

3.1 DESCRIPTION

5.2 DESCRIPTION

6.2 DESCRIPTION

10.2 DESCRIPTION

14.2 DESCRIPTION

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ML19094A132: Difference between revisions

Latest revision as of 21:54, 16 March 2020

Text

SUMMARY

1.1 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

2.2 DESCRIPTION

4.3 DESCRIPTION

5.2 DESCRIPTION

6.2 DESCRIPTION

8.2 DESCRIPTION

9.2 DESCRIPTION

10.2 DESCRIPTION

SUMMARY

2.2 DESCRIPTION

3.1 DESCRIPTION

5.2 DESCRIPTION

6.2 DESCRIPTION

10.2 DESCRIPTION

14.2 DESCRIPTION

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