ML20214L550

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Nonproprietary Amend 1 to RESAR-SP/90 Preliminary Design Approval Module 6/8, Secondary Side Safeguards Sys/Steam & Power Conversion
ML20214L550
Person / Time
Site: 05000601
Issue date: 11/30/1986
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19292G320 List:
References
NUDOCS 8612030113
Download: ML20214L550 (19)


Text

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s WESTINGHOUSE CLASS 3 i

AMENDMENT 1 TO RESAR-SP/90 PDA MODULE 6/8 SECONDARY SIDE SAFEGUARDS SYSTEM / STEAM AND POWER CONVERSION O

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8612030113 861118 i

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WAPWR-SSSS/SPCS AMENDMENT 1 T191e:1d NOVEMBER, 1986

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AMENDMENT 1 TO RESAR-SP/90 PDA MODULE 6/8 SECONDARY SIDE SAFEGUARDS SYSTEM / STEAM AND POWER CONVERSION Instruction Sheet I

Remove current pages 10.3-7/10.3-8, 10.3-9/10.3-10 and replace with revised pages 10.3-7/10.3-8,10.3-9/10.3-10.

O Remove current page 10A.3-1 and replace with revised page 10A.3-1.

Insert remainder of package (page 410-1 through page 410-12) behind Questions / Answers tab.

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' O AMENDMENT 1 WAPWR-SSSS/SPCS NOVEMBER, 1986 4797e:1d

energized which cause the high pressure hydraulic fluid to be dumped to a fluid reservoir.

The redundant electrical solenoids are energized from separate Class 1E sources.

The valves are designed to close in less than 5 seconds against the flows associated with line breaks on either side of the valve, assuming the most limiting normal operating conditions prior to occurrence of the break.

The main steam bypass valve is used when the MSIVs are closed to permit warming of the main steam lines prior to startup.

The bypass valves are air operated globe valves and are normally closed.

For emergency closure, either of two separate solenoids, when de-energized, will result in valve closure.

l Electrical solenoids are energized from a separate Class 1E source.

10.3.2.1.6 Steam Generator Overfill Control Valves i

Each steam generator has a line f rom the upper shell to the EWST (IRC) containing two parallel isolation valves.

These steam generator overfill I

valves are normally closed fast acting, solenoid operated, globe valves.

These 3 inch, carbon steel, [

] pound valves are powered by Class 1E DC +(a c) sources and fail closed on loss of DC power.

The capacity of one valve "is sufficient to prevent steam generator overfill assuming the rupture of one steam generator tube and maximum EFWS flow.

The valves automatically open and close on a high steam generator level signal based on a 2 out of 4 logic.

10.3.2.1.7 Steam Generator Overfill Block Valves Each steam generator has an overfill block valve that is in series with the two overfill control valves.

These overfill block valves are normally open, motor operated gate valves.

These 4 inch, carbon steel, [

] pound valves +(a,c) are powered by Class 1E AC sources.

O APWR-SSSS/SPCS 10.3-7 SEPTEMBER, 1984 1702e:1d

O 10.3.2.2 System Operation NORMAL OPERATION - At low plant power levels, the MSSS typically supplies steam to the steam generator feedwater pump turbines, the auxiliary steam reboiler, and the turbine steam seal system.

At high power levels, these components are typically supplied from turbine extraction steam.

Steam is typically supplied to the second stage steam reheaters in the T-G system when the T-G load exceeds 15 percent.

If a large, rapid reduction in T-G load occurs, steam is bypassed (40 percent of VWO) directly to the condenser via the turbine bypass system.

The system is capable of accepting a 50 percent load rejection without reactor trip and a full ioad rejection without lifting safety valves.

If the turbine bypass system is not available, steam is vented to the atmosphere via the power l

operated relief valves (PORV) and the safety valves, as required.

EMERGENCY OPERATION - In the event that the plant must be shut down and of fsite power is lost, the MSIV and other valves (except to the emergency feedpump turbines) associated with the main steam lines are closed.

The PORV may be employed to remove decay heat and to lower the steam generator pressure to achieve cold shutdown.

If the steam generator PORV for an individual main steam line is not operable, the associated safety valves will provide overpressure protection.

The remaining PORVs are sufficient to achieve cold shutdown.

In the event that a DBA occurs which results in a steam line isolation signal O

(i.e., large steam line break), the MSIV automatically closes.

All the EFW pumps are started automatically on one of several signals (refer to Subsection 1

7. 3.1.1. 6 of RESAR-SP/90 PDA Module 9,

" Instrumentation & Controls and Electric Power," for a description of the logic) with the steam for each O

turbine driven EFW pump automatically supplied f rom one of the SG's.

The closure of three out of four MSIVs will ensure that no more than one steam generator can supply a postulated break.

In addition, closure of the HP turbine steam stop and steam control valves prevents uncontrolled blowdown of APWR-SSSS/SPCS 10.3-8 AMEN 0 MENT 1 NOVEMBER, 1986 1702e:1d l1 --

more than one steam generator following a postulated main steam line break inside the containment.

Coordinated operation of the emergency feedwater system (refer to' Subsection 10.4.9) and PORV or safety valve may be employed to remove decay heat.

10.3.3 Safety Evaluation Safety evaluations are numbered to correspond to the safety design bases of O

Section 10.3.1.1.

SAFETY EVALUATION ONE - The SGIS portion of the MSSS is located in the reactor external building.

This building is designed to withstand the effects of earthquakes, tornadoes, hurricanes, floods, external missiles, and other appropriate natural phenomena.

Sections 3.3, 3.4, 3.5, 3.7 and 3.8 of RESAR-SP/90 PDA Module 7. " Structural / Equipment Design" provide the bases for the adequacy of the structural design of these buildings.

O SAFETY EVALUATION TWO - The SGIS portion of the MSSS is designed to remain functional af ter a SSE.

Subsection 3.7.2 and Section 3.9 of RESAR-SP/90 PDA Module 7, " Structural / Equipment Design" provide the design loading conditions that are considered.

SAFETY EVALUATION THREE - As indicated by Table 10.3-2, no single failure will compromise the system's safety functions.

All vital power can be supplied from either onsite or offsite power systems, as described in Chapter 8.0 of RESAR-SP/90 PDA Module 9, "I&C and Electric Power".

SAFETY EVALUATION FOUR - The MSSS is initially tested consistent with the programs given in Chapter 14.0 of RESAR-SP/90 PDA Module 14, " Initial Test Program" and Chapter 14 of the plant specific applicant's safety analysis O

report.

Periodic inservice functional testing is done in accordance with Subsection 10.3.4.

O APWR-SSSS/SPCS 10.3-9 SEPTEMBER, 1984 1702e:1d

Section 6.6 of RESAR-SP/90 PDA Module 7

" Structural / Equipment Design" provides the ASME Boiler and Pressure Vessel Code,Section XI requirements that are appropriate for the MSSS.

Section 3.2 delineates the quality group SAFETY EVALUATION FIVE classification and seismic category applicable to the SGIS portion of this system and supporting systems.

Table 10.3-1 shows that the components meet the design and fabrication codes given in Section 3.2.

All the power supplies O

and controls necessary for SGIS functions of the MSSS are Class 1E, as described in Chapters 7.0 and 8.0 of RESAR-SP/90 PDA Module 9, "I&C and Electric Power".

SAFETY EVALUATION SIX - Redundant power supplies and power trains operate the MSIVs to isolate safety and nonsafety-related portions of the system.

Branch lines upstream of the MSIV contain normally closed, power-operated relief valves which open and close on steam line pressure.

The atmospheric relief valves fail closed on loss of DC power, and the safety valves provide the overpressure protection.

1 Accidental releases of radioactivity from the MSSS are minimized by the negligible amount of radioactivity in the system under normal operating conditions.

Additionally, the steam generator overfill control valves provide the capability of reducing accidental releases following a steam generator tube rupture.

Detection of radioactive leakage into and out of the system is facilitated by

\\

main steam line radiation monitoring, area radiation monitoring, process radiation monitoring and steam generator blowdown sampling.

SAFETY EVALUATION SEVEN - Each main steam line is provided with safety valves that limit the pressure in the line to preclude overpressurization and remove stored energy.

Each line is provided with a power operated relief valve to permit reduction of the main steam line pressure and remove stored energy to l

achieve an orderly shutdown.

The emergency feedwater system, which is APWR-SSSS/SPCS 10.3-10 SEPTEMBER, 1984 1702e:1d

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10A.3 MAIN STEAM SUPPLY SYSTEM The design of the nonsafety-related portions of the MSSS is the responsibility of the plant specific applicant.

The design must meet the following interface O

criteria:

1.

The applicant's portion of the MSSS piping shall be compatible with the interface connections f rom the SGIS as shown in Figure 10.3-1.

2.

The applicant's portion of the MSSS shall be capable of acconanodating the steam delivered from the SGIS consistent with the parameters given in Table 10.3-1, and with a total pressure drop from the SG to the

)

' turbine stop valves of 25 psi.

3.

The applicant will design the connections to the MSSS between the MSIV's and the turbine stop valves to the following requirements:

Minimize the number and size of these connections and flow capabilities consistent with the Turbine island requirements.

3 Provide a reliable isolation valve in each line and close the valve automatically on the same signal that closes the MSIV.

The i

isolation valve will be a fail closed valve unless that is incompatible with the normal operation of that line.

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l WAPWR-SSSS/SPCS 10 A. 3-1 AMENOMENT 1 T920e:ld NOVEMBER, 1986

J O

3 410.1.

Section 10.3.2.1.1 of RESAR-SP/90 states the following:

"The [ main (10.3) steam] lines are sized for a pressure drop of 25 PSI from the steam (10.A.3) generator to the turbine stop valves..."

Since the plant specific applicant is responsible for designing and supplying all main steam 1

i piping between the steam tunnel / turbine building wall and the turbine stop valves, the piping pressure drop is a design criterion i

needed in design.

The interface section of RESAR-SP/90 (Section 10A.3 and Table 10.3.1) does not provide this information.

Provide this information in those sections.

(SRP 10.3 Part I)

WESTINGHOUSE RESPCNSE:

The plant specific applicant will be designing some, but not all, of 1

the main steam piping; Westinghouse will design all of the piping l

from the SG through the MSIV out to the turbine building just beyond the wall between the turbine building and the main steam penetration l

area (refer to the flow diagram of the SGIS shown on Figure

10. 3-1 ).

However, because the portion of pipe in the turbine building will be dependent on the applicant an interface requirement will be required.

Interface 2 in Subsection 10A.3 has been amended l

l as follows:

2.

The plant specific applicant's portion of the MSSS shall be capable of accomodating the steam delivered from the SGIS consistent with the parameters given in Table 10.3-1, and with a total pressure drop from the SG to the turbine stop O

valves of 25 psi.

410.2 Section 10.3.2.2 of RESAR-SP/90 states the following:

"The closure (10.3) of three out of four MSIVs will ensure that r.o more than one steam O

(10A.3) generator can supply a postulated (main steam line) break.

In addition, closure of the HP turbine steam stop and steam control valves prevents uncontrolled blowdown of more than one steam generator following a postulated main steam line break inside the l

WAPWR-SSSS/SPCS 410-1 AMENDMENT 1 NOVEMBER, 1986 4797e:1d

O containment."

Issue No. 1 of NUREG 0138, " Staff Discussion of Fif teen Technical Issues Listed in Attachment to November 3.1976 Memorandum from Director, NRR to"NRR Staf f" in which credit is taken for all valves downstream of the Main Steam Isolation Valve (MSIV) to limit blowdown of a 'second steam generator in the event of a steam line break' upstream'of,the MSIV is applicable to RESAR-SP/90.

a)

In order to confirm satisfactory performance following such a

~

steam line break provide a tabulation and descriptive text (as appropriate) in the SAR and/or in the interface requirement Section 10A.3.of all potential flow paths that branch off the tr.ain steam lines between the.MSIVs and the turbine stop valves.

For each flow path originating at the main steam lines, provide the following information:

a) System identification b) Maximum allowable steam flow in pounds per hour c) Type of shut-of f valve (s) - desired Maximumallevkblevalve(s) size d) e) Quality of the valve (s) f) Design code of the valve (s) g) Maximum closure time of the valve (s) h) Desired actuation mechanism of the valve (s) (i.e., Solenoid operated, motor operated, air operated diagram valve, etc.)

i) Desired motive or power source for the valve actuating mechanism WAPWR-SSSS/SPCS 410-2 AMEN 0 MENT 1 NOVfMBER, 1986 4797e:ld

O WESTINGH0llSE__ RESPONSE:

(a) The portion of the MSSS that is downstream of the SGIS is not part of the nuclear power block.

As a result the number, size, O.

etc. of each connection to the main steam piping between the MSIV's and the turbine stop valves is dependent on the plant specific applicant's design.

a) The following interface requirement has been added to Subsection 10A.3:

3.

The plant specific applicant will design the connections to the MSSS between the MSIV's and the turbine stop valves to tM following requirements:

Minimize the number and size of these connections and flow capabilities consistent with the Turbine O

island requirements.

Provide a reliable isolation valve in each line and close the valve automatically on the same signal that closes the MSIV.

The isolation valve will be a l

fail closed valve unless that is incompatible with the normal operation of that line.

b)

In the event of the postulated accident, termination of steam O

flow from all systems identified above, except those that can be used for mitigation of the accident, is required to bring the reactor to a safe cold shutdown.

For these systems describe what design features should be incorporated by the plant specific applicant to assure closure of the steam shut-off valve (s).

Describe what operator actions (if any) would be required.

O WAPWR-SSSS/SPCS 410-3 AMENDMENT 1 4797e:ld NOVEMBER, 1986

t O

WESTINGHOUSE RESPONSE:

P (b) Any accident that results in an automatic MSIV closure would not require operator actions to isolate these lines to achieve safe shutdown considering the interface 10A.3.3 listed above.

In the case that the accident did not result in an automatic MSIV closure the operator would follow the appropriate emergency procedure and isolate the connections to the MSSS that should be isolated.

If following isolation of these connections their use was desired then the operator would have to follow the required procedures to unisolate them.

l c) If the systems that can be used for mitigation of the accident are not available or the decision is made to use other means to shut down the reactor describe how these systems are to be secured to assure positive steam shutoff.

Describe what operator actions (if any) are required.

If any of the reauested information is presently included in the SAR text, provide only the references where the information may be found.

(SRP 10.3, Parts II and III):

WESTINGHOUSE RESPONSE:

1 (c) See above item.

O 410.3 Section 10.3.2.2 of RESAR-SP/90 states that on a low-low level in two (10.3) steam generators, steam is automatically provided to each emergency i

(10.4.9) feedwater pump turbine. This statement appears to conflict with the statements given in Section 10.4.9.3 which state that a safety For each injection signal actuates the emergency feedwater system.

of the accident situations described in Section 10.4.9.3 state whether the motor driven and steam driven emergency feedwater pumps are actuated at the same time or at different times.

If the WAPWR-SSSS/SPCS 410-4 AMEN 0 MENT 1 NOVEMBIR, 1986 4797e:1d

actuation times are different, describe the signal and/or condition which activates the pumps, and the time between the motor driven pump's actuation and the steam driven pump's actuation.

Describe the effect on reactor shutdown / system parameters (pressure, J

temperature, steam generator level, etc.) if the pump starting times are different.

(SRP 10.4.9 Part Ill)

WESTINGHOUSE RESPONSE:

Subsection 10.3.2.2 is in error and should state that 'All the EFW pumps are started automatically on one of several signals (refer to Subsection 7.3.1.1.6 of RESAR-SP/90 PDA Module 9, " Instrumentation &

Controls and Electric Power," for a description of the logic) with the steam for each turbine driven EFW pump automatically supplied from one of the SG's."

410.4 In accordance with Standard Review Plan 10.4.7 (April 1984) verify (10.4.7) that the feedwater control valve and controller are designed to be (10A.4.7) stable and compatible with the system-imposed operating conditions (e.g.,

control functions required, range of control and pressure drop characteristics, valve stroke, trim, etc.).

Test data or operating experience data shall be used where available.

In addition provide the necessary interface criteria for the plant operating and maintenance procedures and comit to review those procedures to assure that precautions for avoidance of steam /waterhammer and waterhammer occurrences have been provided or justify not doing so.

(

WESTINGHOUSF RFSPONSE:

Because RESAR-SP/90 is a preliminary design the detailed design of the main feedwater control valves and their control system has not been done at this time.

Refer to the answer to question 410.6 for a discussion of operating and maintenance procedures related to Os waterhamer.

WAPWR -SSSS/SPCS 410-5 AMENDMENT 1 4797e:1d NOVEMBER, 1986

O Provide the necessary interface criteria for the preoperational test 410.5 (10.4.7) program that will verify that unacceptable feedwater hamer will not (10A.4.7) occur using the plant operating procedures for normal and emergency restoration of steam generator water level following loss of normal O

feedwater and possible craining of the feed ring.

(SRP 10.4.7 Part III).

WESTINGHOUSE RESPONSL:

G Preoperational tests will be performed on a RESAR-SP/90 plant to verify that unacceptable feedwater hammer does not occur using the plant operating procedures for normal and emergency restoration of SG water level following loss of normal feedwater.

The interface criteria will be written at a later date when suf ficient desig-detail is available.

410.6 Section 10.4.7.3 Safety Evaluation Nine of RESAR-SP/90 briefly O

describes the features employed in the feedwater system to prevent (10.4.7)

(10A.4.7) flow instabilities due to steam void collapse (i.e. waterhammer) in the main feedwater line.

The features described are those recommended in NUREG-0291 "An Evaluation of PWR Steam Generator Water Hammer."

NUREG-1190 describes a similar type of waterhammer event that occurred in the main feedwater line at San Onofre Unit 1, on November 1, 1985.

Describe how your design, control systems, and operating and maintenance procedures for the main and auxiliary feedwater systems will preclude this event from occurring in RESAR.

Provide in Section 10A.4.7 and 10A.4.9 any interf ace requirements that may be necessary to preclude this event (i.e.

operating procedure guidelines, maintenance criteria, valve design, etc.).

(SRP 10.4.7 Part II)

O l

1 WAPWR-SSSS/SPCS 410-6 AMENDMENT 1 NOVEMBER, 1986 4797e:1d

O WESTINGHOUSF RFSPONSE:

The features incorporated into RESAR-SP/90 to preclude feedwater hamer would have prevented the feedwater hamer event at San Onof re, Unit 1, 11/21/85.

In addition, during the final design of RESAR-SP/90, several additional considerations will be made to further reduce the possibility of a feedwater hamer event in a RESAR-SP/90 plant.

The specific design of the RESAR-SP/90 main feedwater check valves has not been selected at this time, however one consideration for their selection will be a greater assurance that the valve will not come apart inadvertently. Procedures will also be developed to functionally test these check valves on a

periodic basis.

Consideration will also be given to installing a temperature sensor on the main feedwater line inside containment so that back leakage can be detected.

Note that the design of the main feedwater valve, the instrumentation, and the associated procedures are part of the nuclear power block scope and as such will not be addressed in interf ace criteria.

410.7 In accordance with SRP 10.4.9 Part III.3 show that the design of the (10.4.9) emergency and start-up feedwater systems meet the generic recomen-(10A.4.9) dations of NUREG-0611.

Where the systems do not meet the recommendations provide a justification.

In addition, provide in Section 10A.4.9 any interf ace criteria that may be necessary to meet O

the NUREG recomendations.

WESTINGHOUSE RESPONSE:

O The following shows how the EFWS meets the generic recomendations of NUREG-0611.

O WAPWR-SSSS/SPCS 410-7 AMENDMENT 1 NOVEMBER, 1986 4797e:1d

Technical specifications have not been written for RESAR-SP/90; however, it is anticipated that they# will require all four EFW pumps to be available.

The specific test frequency and the outage time window will be evaluated to determine if some relaxation can be made to improve the maintainability of the pumps consistent with the probabilistic risk assessment (PRA).

There is no single valve that can isolate the suction flow to O

more than one of the 4 EFW pumps.

In addition, the manual suction and discharge valves for each pump are locked open.

The EFWS is designed to provide sufficient flow to mitigate a loss of main feedwater flow and reactor trip from 102 % power.

The maximum flow rate that the EFWS can deliver is limited by cavitating orifices, not by throttle valves.

The maximum flow rate of the EFWS will be shown not to cause feedwater hamer.

O The RESAR-SP/90 EFWS has two primary water supplies, one for each sub-system.

Both EFW storage tanks are located inside the auxiliary building and as a result should be available even in the event of external missiles, floods, SSE, etc.

In addition the RESAR-SP/90 is designed for safety grade cold shutdown such that the plant can be transitioned to RHR operation before the EFW storage tanks will empty.

Emergency procedures will be written later for the various possibilities of loss of EFW storage tanks.

O The RESAR-SP/90 turbine driven EFW pumps are independent of AC power; their oil cooling is provided by the pumped fluid, their speed controls are mechanical / hydraulic with all power coming O

from a shaf t driven oil pump, and the steam isolation valve is a f ail open air operated valve. As such these pumps can operate for an indefinite time without AC power.

O WAPWR-SSSS/SPCS 410-8 AMENDMENT 1 NOVEMBER,1986 4797e:1d

O Procedures will be developed for the RESAR-SP/90 EFWS to ensure that its flow path is verified at appropriate tines.

In addition, all remote operated valves in the system have position monitoring lights and alarms for continuous monitoring of the O,

flow path.

The RESAR-SP/90 has saf ety grade, redundant automatic actuation signals.

It also has manual system level actuation from the main control room. The AC powered EFW pumps are sequenced on the emergency diesels.

The RESAR-SP/90 EFWS has redundant level indications and low level alarms on each of the two EFW storage tanks.

In the pump testing in the factory, as well as in the pre-operational plant testing, the EFW pumps will be run for more than 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.

Each SG has an emergency feedwater line with a safety grade flow instrument that provides a safety grade indication in the main control room.

The RESAR-SP/90 EFWS does not require local manual valve re-alignment to perform periodic surveillance tests.

Periodic testing does require the operators to close a motor operated valve in the discharge of the pump being tested.

The position of these valves is monitored and alanned.

Also in the RESAR-SP/90 there are 4 pumps so that when one is being tested there are still 3 others that are aligned for automatic actuation.

O WAPWR-SSSS/SPCS 410-9 AMENDMENT 1 NOVEMBER, 1986 4797e:1d

4 O

Provide the results of any auxiliary feedwater system reliability 410.8 (10.4.9) evaluation that has been performed in accordance with item II.E.1.1.

(10A.4.9) of NUREG-0737 (or justify not doing so).

WESTINGHOUSE RESPONSE:

The reliability of the SFWS and the EFWS have been calculated as a part of the RESAR-SP/90 probabilistic risk assessment (refer to O

Subsection 3.7.2 and Table 3.7.2-2 of RESAR-SP/90 PDA Module 16 (Volume 2), "Probabilistic Safety Study").

The reliability of the EFWS by itself is high relative to current AFWS and together with the SFWS represents a significant improvement.

The following represents the SFWS and the EFWS reliabilities:

LMFW Loss all AC

+(a,c)

SFWS EFWS SFWS & EFWS 410.9 Provide the source of power for the motor driven emergency feedwater (10.4.9) pump and the motor-operated valves. (SRP 10.4.9 Part III)

WESTINGHOUSE RESPONSE:

The EFW pump number 1 is a motor driven pump that is located in the "A"

saf eguards area and is powered by train "A";

its associated motor operated valve is also powered by train "A".

EFW pump number 2 is a motor driven pump that is located in the "B" safeguards area and is powered by train "B"; its associated motored operated valve O

is powered by train "B".

EFW pump number 4 is a turbine driven pump that is located in the "A" safeguards area; its associated motor operated valve is powered O

by batteries associated with train "A".

EFW pump number 3 is a WAPWR-SSSS/SPCS 410-10 AMENDMENT 1 NOVEMBER, 1986 4797e:1d

h turbine driven pump that is located in the "B" safeguards area; its associated motor operated valve is powered by batteries associated with train "B".

Inside containment there is a cross connection between EFW pumps 1 &

I 4 that has two MOV's.

One of these valves is powered by train "A" and the other valve is powered by train

'B'.

The same applies to the cross connection between EFW pumps 2 & 3.

O 410.10 Section 10.4.9.2.2.2 of RESAR-SP/90 states that "in the event the (10.4.7) SFWS [startup feedwater system] fails to function properly following l

(10A.4.9) a plant transient...the EFWS is automatically started."

Describe the controls, sensors and/or indicators which automatically activate the emergency feedwater system for this event.

Provide the necessary interface criteria in Section 10A.4.10.

(SRP 10.4.9 Part III) i WESTINGHOUSE RESPONSE:

The control to start the EFWS on failure of the SFWS is described in Subsection 7.3.1.1.6 of RESAR-SP/90 PDA Module 9, " Instrumentation &

l Controls and Electric Power" and shown on the functional diagrams (Figure 7.2-1 sheet 8).

The EFWS is actuated on a Low-1 SG 1evel signal f rom the narrow range in any SG coincident with a low SFWS flow to any SG. The SFWS flow instruments are part of the nuclear power block and are shown on Figure 10.3-1 of RESAR-SP/90 PDA Module 6/8, " Secondary Side Safeguards System / Steam and Power Conversion."

Because all of the controls and instruments are part of the nuclear power block, no interface requirements are required.

410.11 The interf ace criteria in Section 10A.4.9 of RESAR SP/90 states that (10A.4.9) the plant specific applicant needs to provide an alternate water source capable of maintaining the plant in hot standby conditions with one reactor coolant pump operating for two days after a reactor O

WAPWR-SSSS/SPCS 410-11 AMENDMENT 1 NOVEMBER, 1986 4797e:1d

trip.

The amount of water necessary to accomplish this task is O

described as " sufficient."

Provide the minimum amount of water necessary to accomplish this task.

(SRP 10.4.9 Part III).

WESTINGHOUSE RESPONSE:

The total volume of water required to maintain the plant in a hot standby condition with one RCP operating for two days following a reactor trip is [

] gal.

Of this total there is already +(a c)

O

] gal in the EFW storage tanks, so that the difference of +(a,c)

[

l

[

] gal should be contained in the plant specific applicant's +(a c) tanks.

Note.that the plant specific applicant's water volume is not safety grade.

O O

O O

WAPWR-SSSS/SPCS 410-12 AMENDMENT 1 NOVEMBER, 1986 4797e:ld

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _