ML20199L172

Nonproprietary Amend 1 to Westinghouse Advanced PWR RESAR-SP/90 Preliminary Design Approval Module 1, Primary Side Safeguards Sys

ML20199L172
Person / Time
Site:	05000601
Issue date:	05/30/1986
From:	WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19292F556	List: ML20199L163 ML20199L172 NRC-86-3138, Forwards Proprietary & Nonproprietary Amend 1 to Westinghouse Advanced PWR RESAR-SP/90 Preliminary Design Approval Module 1, Primary Side Safeguards Sys. Proprietary Amend Withheld (Ref 10CFR2.790)
References
NUDOCS 8607090329
Download: ML20199L172 (257)

Text

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1 WE5TINGHOUSE CLHSE 3 I

AMENDMENT 1 TO kESAR-SP/90 PDA MODULE 1 "PRIMAEY SIDE SAFEGU4RDS SYSTEM" l

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l 8607090329 860609 PDR ADOCK 05000601 K PDR wats -vEd: AMENDMEt.1 1 MAi, 195::

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i AMENDMEN1 1 TO RE54k-SP/90 PDA MODULE f " PRIMARY SIDE SAFEGUARD 5 SYSTEM" l

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! INSTRUCTIUN SHEET l l

1. Replace current lable 1.3-1, Sheet 1 (page 1.3-2) with revised Table 1.3-1, Sheet 1 (page 1.3-2).

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i 2. Replace current page 5.4-1 with revised page 5.4-1.  !

I i 3. Replace current page 6.3-14 with revised page 6.3-14.

4. Replace current page 6.3-16 with revisec page 6.3-16. ,

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5. keplace current page o.3-20 witn revised paae 6.3-20.  ;

I Feplace current page 6.3-31 witn revised pace 6.3-31.

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7. Replace current pace o.3-32 with revised page 6.3-32. j

8. Peplace current page e.3-35 with revised page 6.3-35.

9 heplace current page 6.3-36 with revised page 6.3-36.

10. Feplace current pace 6.3-38 with revised pace 6.5-38.

11. Feplace current pace 6.3-42 with revised page 6.3-42.  ;

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12. keplace current pace o.3-As with revised page 6.3-49.

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13. Feplace current page 15.5-1 with revised page 15.5-1.

14. Replace current page 15.5-2 witn revised page 15.5-2.

1 1 15. Replace current pace 15.6-2 with revised page 15.6-2.

16. Replace current page 15.6-12 with revised page 15.6-12.

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17. Replace current Table 15.6.4-1 (page 15.6-21) with revised Table 15.6.4-1 (page 15.6-21) t

18. Peplace current Figure 15.6.4-33 with revised Figure 15.6.4-33, )

i 19, insert remainder of pac 6 age (page 440-1 through Figure 440.2398) behind Questions / Answers tab.

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O AMENDMENT 1 WAPL'-E555 2 MAY, 198' i

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TABLE 1.3-1 (Sheet 1 of 2)

DESIGN COMPARISON O Parameter or Feature RESAR-SP/90 RESAR-414 RESAR-3S RESAR-41 Residual heat removal ~ 400 l1

1. Initiation pressure (psig) - 400 ~ 425 ~ 425

2. Initiation / completion ~350/150 ~350/140 ~350/140 ~350/140 0 temperature (*F)

3. Component cooling water 105 105 95 105 design temperature (*F)

4. Cooldown time after - 16 - 16 - 16 ~8 initiation (hr) )(a,c) 46.5 37.4 39.4 l1

5. Heat exchanger removal [

(106 Btu /hr) (4 provided)

Accumulators 4 4 4 4

1. Number

2. Operating pressure, normal 600 700 600 600 l1 (psig)

3. Nominal operating water )(a,c) 1,400 850 1,500

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volume, each (ft 3) l

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1.3-2 AMEN 0 MENT 1 O WAPWR-PSSS 4854e:ld MAY, 1986

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5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS

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d 5.4 COMPONENT AND SUBSYSTEM DESIGN 5.4.7 Residual Heat Removal System The f unctions performed by a conventional nuclear plant residual heat removal (RHR) system are integrated within the integrated safeguards system (ISS) of the WAPWR. Those components within the ISS that perform an RHR function are O the four low head pumps, four RHR heat exchangers, and the associated valves, piping, and instrumentation.

Following initial cooldown of the plant by steam dump to the main condenser, the four low head pumps would be aligned to recirculate reactor coolant through the core by taking suction from the RCS hot legs and returning the coolant to the reactor vessel through the four RHR heat exchangers. The four RHR pumps and four heat exchangers would be capable of reducing the RCS coolant temperature from 350*F to 150*F in less than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> following reactor shutdown. It should be emphasized that the ISS provides four RHR subsystems which would permit three of four subsystems to be taken out of service during long-term shutdown operation.

The low head pumps have multiple uses. During plant cooldown and refueling operations, they act as conventional RHR pumps. In their accident mitigation role, they act as containment spray pumps (see Section 6.2.2) or could be aligned to provide a long term ECCS function (see Section 6.3.2.2.2).

Additionally, the pumps are used to transfer refueling water from the ref ueling canal and the Emergency Water Storage Tank (EWST) at the beginning of ref ueling operations. Refueling water is returned to the EWST from the refueling cavity at the end of refueling operations by a gravity drain.

The RHR heat exchangers perform the heat removal function during plant

' cooldown and refueling operations as well as during accident recovery operations (see Section 6.3.2.2.6). ]1

5. 4 -1 AMENDMENT 1 MAPWR-PSSS MAY, 1986 4854e:ld

' Water can be removed f rom the core reflood tanks by opening the appropriate valves in the test line and permitting flow to return to the EWST. Periodic checks of the core reflood tank boron concentration are made through the sampling system.

O Redundant level and pressure indicators are provided on each core reflood tank with readouts on the main control board. Each indicator is equipped with high and low-level alarms. The margin between the minimum operating pressure or level and the maximum operating pressure or level provides a range of O acceptable operating conditions. The band width is sufficient to minimize the f requency of adjustments in the core reflood tank (CRT) pressure or level 1 required to compensate for leakage.

The design parameters for the core reflood tanks are provided in Table 6.3-2.

6.3.2.2.5 Emeraency Water Storace Tank (EWST)

The EWST is located at the lower elevation inside the containment building and 3

provides a continuous suction source for the high and low head pumps, thereby eliminating the switchover f rom injection to recirculation. The EWST and loop compartments would be arranged to minimize the containment cleanup in case of minor accidents, such as reactor coolant pump seal failure or instrument line breaks. Any discharge f rom the pressurizer relief tank should also be routed to this tank. The required water volume depends on the Refueling Canal Volume which is not expected to exceed [ ] gallons. (a.cl Analyses are performed to determine the minimum water level in the EWST during recirculation. These analyses consider the amount of water trapped in lower containment compartments and the delay time f or water to return to the EWST.

The design parameters for the EWST are provided in Table 6.3-2.

O 6.3.2.2.6 Residual Heat Removal (RHR) Heat Exchangers Four RHR heat exchangers are provided, with one RHR heat exchanger assigned to

'I each of the four subsystems. Each exchanger is sized to remove [ ] percent 6.3-14 AMENDMENT 1 WAPWR-PSSS MAY, 1986 4854e:ld i

a. Where possible, packless valves are used,

b. Other valves which are nonnally open, except check valves and those which perform a control function, are provided with backseats to limit stem leakage.

c. Normally closed globe valves are installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (radioactive) water.

d. Relief valves are enclosed, i.e., they are provided with a closed bonnet.

Motor-Operated Gate Valves The seating design of all motor-operated gate valves is of the crane flexible wedge design. These designs release the mechanical holding force during the first increment of travel so that the motor operator works only against the f rictional component of the hydraulic unbalance on the disc and the packing box friction. The discs are guided throughout the full disc travel to prevent chattering and to provide ease of gate movement. The seating surfaces are hard faced to prevent galling and to reduce wear.

Where a gasket is employed for the body to bonnet joint, it is either a fully trapped, controlled compression, spiral wound asbestos gasket with provisions for seal welding, or it is of the pressure seal design with provisions for seal welding. The valve stuffing boxes are designed with a lantern ring leakoff connection with a minimum of a full set of packing below the lantern ring and a minimum of one-half of a set of packing above the lantern ring. A full set of packing is defined as a depth of packing equal to 1-1/2 times the stem diameter.

l The motor operator incorporates a " hammer blow" feature that allows the motor to l

impact the discs away from the backseat upon opening or closing. This " hammer N blow" feature not only impacts the disc but allows the motor to attain its l operational speed prior to impact. Valves which must function against system '

pressure are designed such that they function with a pressure drop equal to full 1 system pressure the valve disc. L l AMENDMENT 1 l

WAPWR-PSSS 6.3-16 MAY, 1986 4854e:ld

recirculation or operator error and maintain the performance objectives desired in subsection 6.3.1. Separate trains of pumps, heat exchangers and flow paths are provided for redundancy. The initiating signals for the ECCS are derived from independent sources as measured from process (e.g, pressurizer low pressure) or environmental (e.g, containment high pressure) variables. Redundant as well as functionally independent variables are measured to initiate the Safeguards signal. Each train is physically separated and protected where necessary so that a single event cannot initiate a common failure. Power sources are divided into independent trains supplied from the separate emergency buses supplied f rom of f site power. Sufficient diesel generating capacity is maintained on site to provide required power to each train. The diesel generators and their auxiliary systems are completely independent and dedicated to one or two of the trains.

The preoperational testing program ensures that systems, as designed and 1 constructed, will meet the functional requirements. The ECCS is designed with the ability for on-line testing of most components so the availability and operational status can be readily determined. In addition to the above, the integrity of the ECCS is ensured through examination of critical components during the routine inservice inspection.

1 The reliability program extends to the procurement of ECCS components such that only designs which have been proven by past use in similar applications are acceptable for use. The quality assurance program as described in Chapter 17 ensures receipt of components only af ter manufacture and test to the applicable codes and standards.

6.3.2.5.1 Active Failure Criteria The failure of a powered component, such as a piece of mechanical equipment,'

component of the electrical supply system or instrumentation and control equipment, to act on command to perform its design function is considered an active failure. Examples include the failure of a motor-operated valve to move to its correct position, the failure of an electrical breaker or relay to l

respond, the f ailure of a pump, fan, or diesel generator to start, etc.

l WAPWR-PSSS 6.3-20 AMEN 0 MENT 1 4854e:ld MAY, 1986

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a. The calculated peak fuel element clad temperature is less than 2,200'F

b. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

c. The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The cladding oxidation limits O, of 17 percent are not exceeded during or after quenching.

d. The core remains amenable to cooling during and after the break.

e. The core temperature is maintained at an acceptably low value and 1 decay heat is removed for an extended period of time, as required by O the longlived radioactivity remaining in t M core.

6.3.3.4 Major Secondary System Pipe Failure Discussion The steam release from a rupture of a main steam pipe would result in an increase in energy removal from the RCS causing a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an insertion of positive reactivity.

There is an increased possibility that the core will become critical and return to power. A return to power following a steam pipe rupture is a potential problem; however, analysis demonstrates that the core is ultimately shut down by the injected boric acid.

Minimum capability for injection of boric acid (2,500 gpm) solution is assumed corresponding to the most restrictive single failure in the ECCS system. For the cases where offsite power is assumed to be available, the high head pumps are assumed to start immediately upon receipt of the "S" signal and to achieve O full speed in five seconds. The water initially within the high head pump piping is assumed to be swept into the RCS (with no credit taken for its boron) before the 2,500 ppm water f rcm the EWST reaches the core. For the cases where offsite power is assumed not to be available, an additional 10 second delay is assumed to start the diesels. The necessary ECCS equipment are then loaded onto the diesels according to the sequencer.

6.3-31 AMENDMENT 1 WAPWR-PSSS MAY, 1986 4854e:1d

N See the " Secondary Side Safeguards" module for the results and conclusions of this analysis.

6.3.3.5 Steam Generator Tube Failure The accident examined is the complete severance of a single steam generator 5

tube at power.

• l Assuming normal operation of the various plant control systems, the following l sequence of events is initiated by a tube rupture: i

a. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side, there is a steam flow /feedwater flow mismatch before the trip as the feedwater flow to the affected steam generator is reduced due to the additional break flow which is now being supplied to that unit.

b. The steam generator blowdown liquid monitor and the condenser of fgas radiaton monitor will alarm, indicating a sharp increase in radioactivity in the secondary system and will automatically terminate steam generator blowdown.

c. Continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure or low DNBR reactor trip. The "S" signal automatically terminates normal feedwater supply ,

and initiates emergency feedwater addition. After reactor trip, the break flow reaches equilibrium at the point where incoming safety

- injection flow is balanced by outgoing break flow. The resultant break flow persists f rom plant trip until operator action is taken to bring the primary system and the f aulted steam generator secondary system pressures into equilibrium.

d. The reactor trip automatically trips the turbine and, if offsite power is available, the steam dump valves open permitting steam dump to the condenser. In the event of a coincident station blackout, the steam O- WAPWR-PSSS 4854e:1d 6.3-32 AMENDMENT 1 MAY, 1986

An "S" signal normally results in a reactor trip followed by a turbine trip.

However, it cannot be assumed that any single f ault that actuates ECCS will also provide a reactor trip. If a reactor trip is generated by as spurious signal, the operator should determine if the signal was transient or steady state in nature and if the safety injection signal must be blocked. For a O spurious occurrence, the operator would terminate the safety injection and maintain the plant in the hot shutdown condition. If the SIS actuation 1 instrumentation future plant operation would be in accordance with the Technical Specifications.

Conclusions Results of the analysis show that spurious ECCS operation without immediate reactor trip presents no hazard to the integrity of the RCS.

6.3.4 Tests and Inspections 6.3.4.1 ECCS Performance Tests Preliminary operational testing of the ECCS can be conducted following flushing and hydrostatic testing, with the system cold and the reactor vessel 1 head removed. Subsequent system performance testing can be conducted during each major fuel reloading operation with each subsystem aligned to take suction f rom the EWST and to deliver to the EWST via the system test line.

Each pump can also inject into the reactor vessel, with the overflow f rom the reactor vessel spilling into the refueling canal. Simultaneously, the safety injection block switch is reset and the breakers on the lines supplying off site power are tripped manually so that operation of the emergency diesels is tested in conjection with the ECCS. This test should provide information including the following facets:

a. Satisfactory safety injection "S" signal generation and transmission.

I Proper operation of the emergency diesel generators, including

b.

sequential load pickup.

I 6.3-35 AMENDMENT 1 s WAPWR-PSSS MAY, 1986 4854e:ld i

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c. Pump starting times.

d. Pump delivery rates.

Separate flow tests of the low head and high head pumps should be conducted

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\ during any system performance test operation to verify the pump head / flow characteristics. In addition, these tests are required to establish / verify flows in conjunction with the required pump discharge flow rates for both the s reactor vessel injection and hot leg injection modes of operation. During

\ these tests, the pumps are aligned to take suction f rom the EWST and to discharge into the reactor vessel through the injection lines. More specifically, the system performance tests are required to ensure that the appropriate sized flow-restricting orifice plates are installed in the high head pump miniflow lines and the high head and low head pump discharge headers.

Each accumulator and core reflood tank is filled with water from the EWST and l1 pressurized with nitrogen with the motor-operated valve on the discharge line closed. The valve is opened and the accumulator and core reflood tank allowed to discharge into the reactor vessel as part of the operational startup testing with the reactor vessel head off.

6.3.4.2 Reliability Tests and Inspections Routine periodic testing of ECCS components and all necessary support systems at power is planned. Valves which operate af ter a LOCA are operated through a l complete cycle, and pumps are operated individually in this test on their miniflow lines.

O If such testing indicates a need for corrective maintenance, the redundancy of equipment in these systems permits such maintenance to be performed without shutting down or reducing load under certain conditions. These conditions include considerations such as a period within which the component should be

\ restored to service and the capability of the remaining equipment to provide the minimum required level of performance during such a period.

6.3-36 AMENDMENT 1 O

WAPWR-PSSS 4854e:ld MAY, 1986

The signals that are generated by the protection logic and used to initiate O the "S" signal are the following;

a. Pressurizer trip signal, produced by two-out-of-four (2/4) pressurizer low-pressure signals  ;

b. Hi-1 containment pressure trip signal, produced by two-out-of-four (2/4) containment Hi-1 pressure signals

c. Steam line low-pressure signal, produced by two-out-of-four (2/4) steam line low-pressure signals in one line

d. Excessive cooldown, produced by low T-cold signals in two-out-of-four (2/4) loops coincident with a neutron flux of below 10 percent

e. Manual safety injection actuation from the control board The actuation signal that initiates containment isolation phase A and containment ventilation isolation is referred to as the "T" signal. The "T" signal is initiated from the same protection logic signals that produced the "S" signal, except that a separate manual actuation switch is provided on the control board that permits the operator to initiate containment isolation phase A actuation without initiating the ECCS. In addition, the "S" signal reset is separate f rom the "T" signal reset.

The actuation signal that initiates spray actuation and containment isolation phase B is referred to as the "P" signal. The signals that are generated by the protection logic and used to initiate the "P" signal are the following:

a. Hi-3 containment pressure trip signal, produced by two-out-of-four (2/4) containment Hi-3 pressure signals A b. Manual actuation from control board 6.3-38 AMENDMENT 1 WAPWR-PSSS MAY, 1986 4854e:1d

b. Low Head Pump Discharge Header Temperature (TE-912, 913, 914, and 915)

There is one temperature element in the discharge header of ea:h low head pump with readout on the main control board. These temperature trans-mitters represent the inlet temperatures to each RHR heat exchanger and they are recorded, in conjuntion with the RHR heat exchanger outlet tem-perature (TE-924, 925, 926, and 927), by a dual-point recorder on the main control board to indicate the delta temperature reduction of the RHR flow.

These temperature elements are also used to provide input to temperature channel bistables that are part of the protection logic used to ensure that component cooling water flow is initiated to the corresponding RHR heat exchangers. The automatic opening of the RHR heat exchanger /

component cooling water isolation valves would be initiated in the event that actuation signals were generated by the RHR pump discharge temperature logic. A temperature actuation signal would be generated when l 1 a single temperature channel bistable receives a temperature signal from a corresponding temperature element, higher than a pre-determined temperature setpoint.

c. RHR Heat Exchanger Outlet Temperature (TE-924, 925, 926, and 927)

There is a temperature element in the outlet of each RHR heat exchanger downstream of the flow bypass return. Readout is on the main control board. The temperature is recorded in conjunction with the RHR heat exchanger inlet temperature (TE-912, 913, 914, and 915) on the main control board by a dual-point recorder to indicate the delta temperature l reduction of RHR flow.

d. RHR Heat Exchanger Outlet Temperature - Local (TI-916, 917, 918, and 919)

There is a temperature indicator in the outlet of each RHR heat exchanger. It provides a means of performance verification and heat balance when used in conjunction with the RHR heat exchanger inlet temperature indication of TE-912, 913, 914, 915.

O WAPWR-PSSS 4854e:ld 6.3-42 AMENDMENT 1 MAY,1986

TABLE 6.3-2 (Sheet 2 of 6)

INTEGRATED SAFEGUARDS SYSTEM COMPONENT PARAMETERS Low Head Pumps lg Number 4 Type Horizontal or Vertical O Runout Flow Rate, gpm Runout Head, ft ta,c)

Design Flow Rate, gpm Design Head, ft Shutoff Head, ft Miniflow, gpm ,

Motor Capacity, bhp Speed, rpm Discharge Design Pressure, psig Suction Design Pressure, psig Design Temperature, 'F NPSH Required, at pump suction (ft)

NPSH Available at pump suction (ft)

Seismic Category I l

Design Code ASME III, Class 2 O

II} Based on a horizontal pump.

O WAPWR-PSSS 6.3-49 AMENDMENT 1 O 4B54e:ld MAY, 1986

15.5 INCREASE IN REACTOR COOLANT INVENTORY Discussion and analysis of the following event is presented in this section:

1. Inadvertent operation of the emergency core cooling system during power operation This event, considered to be ANS Condition II, can result in an increase in reactor coolant inventory, if the reactor coolant pressure falls below the 1 shut-off head of the safety injection pumps following a reactor trip.

An increase in the reactor coolant inventory due to CVCS malfunctions will be discussed in the " Reactor Coolant System" module.

15.5.1 Inadvertent Operation of the Emeraency Core Coolina System Durira Power Operation

! 15.5.1.1 Identification of Causes and Accident DescriDtion Spurious emergency core cooling system (ECCS) operation at power could be caused by operator error or a false electrical actuation signal. A spurious signal may originate from any of the safety injection actuation channels as described in Section 7.3 of the "I&C and Electric Power" module.

Following the actuation signal, the high head safety injection pumps will start automatically. However, since the shutoff head of the safety injection l pumps is less than the reactor coolant system (RCS) pressure, no flow will be delivered to the RCS. The passive accumulator injection system and the passive core reflood tanks also provide no flow st normal RCS pressure.

A safety injection system (SIS) signal normally results in a reactor trip followed by a turbine trip. However, it cannot be assumed that any single fault that actuates the SIS will also produce a reactor trip. If a reactor trip is generated by the spurious SIS signal, the operator should determine WAPWR-PSSS 15.5-1 AMENDMENT 1 4854e:ld MAY, 1986

if the spurious signal was transient or steady state in nature and if the

' O' safety injection signal must be blocked. Following reactor trip the safety injection system will not deliver any flow until the RCS pressure falls below the shutof f head of the safety injection pumps. For a spurious occurrence, the operator would stop the safety injection and maintain the plant in the hot s standby condition. If the SIS actuation instrumentation must be repaired, in accordance with the Technical future plant operation would be Specifications.

15.5.1.2 Conclusions This transient is not analyzed since inadvertent operation of the emergency core cooling system without reactor trip does not result in the injection of any borated water since the shutoff head of the safety injection system is well below normal operating pressures. If the spurious safety injection signal produces protection system actuation, the incident simply results in a reactor trip.

O a

O 15.5-2 AMENDMENT 1 WAPWR-PSSS MAY, 1986 1 O 4854e:ld 1

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b) The amount of fuel element cladding that reacts chemically with water or steam does not exceed one percent of the total amount of Zircaloy in the fuel rods in the reactor. l1 c) The clad temperature transient is terminated at a time when the core s geometry is still amenable to cooling. The localized cladding oxidation limits of 17 percent are not exceeded during or after quenching.

d) The core remains amenable to cooling during and af ter the break.

O e) The core temperature is maintained at an acceptably low value and decay heat is removed for an extended period of time, as required by the longlived radioactivity remaining in the core.

These criteria were established to provide significant margin in ECCS performance following a LOCA. Reference 15.6.4-2 presents a recent study in the probability of occurrence of RCS pipe ruptures.

In all cases, small breaks (less than 1.0 ft 2) yield results with more margin to the acceptance criteria limits than large break.

15.6.4.2 Seauence of Events and Systems Operations Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip signal subsequently occurs when A safety injection the pressurizer low pressure trip setpoint is reached.

actuation signal is generated when the appropriate setpoint is reached. These countermeasures limit the consequences of the accident in two ways:

a) Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

WAPWR-PSSS 15.6-2 AMENDMENT 1 MAY, 1986

( 4854e:ld

The results show that the large break LOCA transient is characterized by three d distinct phases: (1) initial heat-up phase due to the core stored energy and the relatively poor positive core flow period leading to the initial clad temperature rise, (2) blowdown cooling phase characterized by influx of water p into the core from the large upper plenum and upper head volumes resulting in a significant blowdown cooling ef fect, and (3) reflood phase with relatively high core velocities driven by water elevation head in vessel downcomer.

l The maximum clad temperature calculated for a large break is [ ] which is (*'CI q

V less than the acceptance criteria limit of 2200*F of 10CFR 50.46. The maximum local metal water reaction is less than 1.0 percent which is well below the embrittlement limit of 17 percent as required by 10CFR 50.46. The total core metal-water reaction is less than 0.3 percent for all breaks, as compared with the 1 percent criterion of 10 CRF 50.46, and the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. No fuel rod burst occurs. As a result, the core temperature will continue to l1 drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.

Small Break Results As noted previously, the calculated peak clad temperature resulting f rom a small break LOCA is less than that calculated for a large break. The limiting small break was found to be less than a 10 in. diameter rupture of the RCS cold leg. A range of small break analyses are presented which establishes the limiting small break. The results of these analyses are summarized in Tables 15.6.4-4 and 15.6.4-5.

O Figures 15.6.4-31 through 15.6.4-39 present the principal parameters of interest for the small break ECCS analyses. For all cases analyzed the following transient parameters are presented:

O a) RCS pressure b) Core mixture height, 15.6-12 AMEN 0 MENT 1 MAPWR-PSSS MAY, 1986 4854e:ld l - -

O TABLE 15.6.4-1 O INPUT PARAMETERS USED IN THE ECCS ANALYSISO (a,c)

O Core Power

• Peak linear power (includes 102% factor Total peaking factor F g Power shape large break-chopped cosine Small break-see Figure 15.6.4-39 Full assembly array Accumulator water volume (minimum) 1 Accumulator tank volume Accumulator gas pressure (minimum)

O Core reflood tank water volume (minimum) l1 Core reflood tank volume Core reflood tank gas pressure (minimum) 11 Safety injection pumped flow Initial loop flow Vessel inlet temperature Vessel outlet temperature Reactor coolant pressure Steam pressure Steam generator tube plugging level

• 2% is added to this power level to account for calorimetric error.

)

O WAPWR-PSSS 15.6-21 AMENDMENT 1 MAY, 1986 2854e:

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' " Questions / Answers" i,

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W APWi -PSss AMENDMENT 1 May, 1986 E

I F ow--rw-.-----m,,

440.30 NUREG-0737, Item III.D.l .1, " INTEGRITY OF SYSTEMS OUTSIDE CONTAINMENT 5.4.7 LIKELY TO CONTAIN RADIDACTIVE MATERIAL FOR PRESSURIZED-WATER REACTORS AND SOILING-WATER REACTORS," contains the statement: " Applicants shall implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids during a serious transient or accident to as-low-as-practical icvels." It also states: " Implement all practical leak reduction measures for all systems that could carry radioactive fluid outside of containment." It continues with discussion of leak reduction and leak determination programs and, with respect to design, contains the statement: "Should consider program to reduce leakage potential release paths due to design ... deficiencies ...." This concern is reflected in 10CFR 50.34(f)(2)(xxvi) which is a requirement for i

leakage control and detection in the design of systems outside containment that contain radioactive materials following an accident. This regulation contains the statement: "The goal is to minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency (1I1.D.1.1)". The applicable section of NUREG-0737 is also referenced in the SRP in review of Section 5.4.7, and the RHR system is specifically identified in the NUREG as one of the systems to which this concern is applicable.

Further clarification of the intent of the referenced wording is provided in NUREG-0660, Task III.D.1 A. which states: "0BJECTIVE:

Perform evaluations to establish additional design features that should be included in the rulemaking proceeding of Item II.B.8. The purpose of these evaluations is to identify features that will reduce l

I the potential for exposure to workers at nuclear power plants and to of fsite populations following an accident." Finally, the referenced Item II.B.8 contains: "NRC will conduct a rulemaking . . . to solicit comments on the issues and facts relating to design features O

necessary to deal effectively with degraded-core and core-melt accidents and to mitigate the consequences. Specific areas ... will include...the characteristics and effectiveness, of possible design features to cope with and mitigate the consequences of these types of WAPWR-PSSS 440-1 AMENDMENT 1 2855e:ld MAY, 1986

accidents; additional and supplemental means of preventing core damage or core-melt accidents through improved engineered safety features; the probabilities and consequences of the various sequences of events that could cause the release of significant amounts of radioactivity to the environment, the expected effectiveness and performance of suggested means to reduce the consequences of such events ....

" While the draf t NUREG-1070 withdraws the rulemaking, the intent of the above sited documents remains, as identified in NUREG-1070.

Under core damage conditions, particulate matter from a damaged core or containment debris might damage pump seals and bearings and valves associated with cooling under post-accident conditions. An obvious advantage of the SP/90 configuration with four RHR subsystems is the potential that less than four are needed for cooling duty. In light of the concerns raised in the above regulations and guidance, please:

a. Address the considerations that have been provided to cope with such conditions.

l l b. Address what other considerations have been provided with respect to equipment operation, if any.

l l RESPONSE:

The present design provides four enclosures (one for each of the SI subsystems) which, although not designed to withstand pressure and return high leakage flow rates to the containment, do provide control of release of radioactive material via atmospheric control and control of leakage via sump pumps. The atmosphere of the enclosures is filtered and then vented. Leakage should be minimized through operating practice.

O 440.31 Assessments of existing PWRs and of those under construction show an 5.4.7 interfacing systems LOCA to be a significant contributor to risk at many plants. The RHR system is described as consisting of four O

WAPWR-PSSS 440-2 AMENDMENT 1 2855e:ld

• MAY, 1986

subsystems which are essentially identical. The description is not '

clear as to whether all portions of these subsystems are located so that LOCA outside of containment (event "V") is effectively eliminated. Therefore, with consideration of these comments and the concerns raised in 440.30:

a. Please discuss the location of all piping, including valves, that are outside of containment and the measures that have been taken so that event "V" is addressed in the SP/90 plant. If certain piping is located within a structure that effectively serves as containment, but is not designed to " safety-related" containment criteria, please indicate such structure and explain its function in control of event "V".

b. If any of the piping described above is not within a structure that effectively provides an ECCS recirculation path in the event of an inter-systems LOCA, please address how the probability of

' the event has been changed relative to existing plants. Of particular concern is the larger number of pipes which lead from the containment to components that are outside of the containment.

c. What consideration was given to designing the entire system to meet RCS operating conditions, thereby decreasing the probability of an inter-systems LOCA and providing a high pressure decay heat removal mode.

O RESPONSE:

a. There is no structure in the SP/90 plant that effectively serves as a containment structure outside of the actual containment.

The CPPE's are not included in the design submitted to NRC for ON review. The piping and valves are generally identified in the SAR material submitted for review. Insofar as event "V" is concerned, the SP/90 design is simpler than the design in O

WAPWR-PSSS 440-3 AMENDMENT 1 285Se:ld MAY, 1986

r existing nuclear plants and low pressure piping is protected by l

i multiple in-series valves and interlocks.

b. A larger number of RHR and SI systems does not necessarily lead to a greater concern with event "V". The configuration is such g

that there are fewer valves outside containment, much less piping W outside containment, and the piping is generally of a smaller diameter. There is considerably less welding than in present comparable plants.

e. This is essentially an economic issue. A design to meet full RCS operating conditions requires that both temperature and pressure conditions be met. This applies to piping, valves, and pumps; and particularly the seals of the latter, where temperature would be of concern. The proposed design is satisfactory; particularly in light of the safeguards built into the design to prevent overpressurization of the RHR system, as discussed in the response to the prior question.

440.32 To provide further information pertinent to the concerns raised in 5.4.7 440.30 and the CP rule (typically items (1)(i), (2)(ii), (2)(v), and (2)(xxvi), as well as to further staff understanding of overall SP/90 response to accident conditions, please provide the following information related to pipe breaks or leaks in lines outside containment associated with the RHR system:

l

a. The design criteria of the pump house with respect to i.bility to withstand a break or leak from any component in systems outside containment, but inside the pump house, that are used for core cooling. Also please provide the conditions of the breaks which are covered by the design criteria, identify any breaks which are not covered by the criteria, and provide the expected failure conditions for the pump house.

O WAPWR-PSSS 440-4 AMENDMENT 1 2855e:1d MAY, 1986

b. The time available prior to core damage and prior to pump house failure and an estimate of core damage status at the time of pump house failure.

c. Describe the alarms available to alert the operator to the presence of a break event outside containment (in the pump house), the recovery approach that will be used by the operator, and the time available for this action.

O RESPONSE:

a. The pump houses in the submitted SP/90 design are not designed to contain breaks. Leaks that are within the capability of the sump pump system can be controlled since the liquid can be removed by the sump pumps. Atmospheric control of radioactive material is provided, as previously discussed,

b. Not applicable since the CPPE design is not part of the SP/90.

O c. This information will be provided in RESAR-SP/90 PDA Module 15 "ACR/ Human Factors."

l 440.33 Please discuss the rationale for EWST size (or reference the section i 5.4.7 which will contain this discussion). Include consideration of the decrease in water level which may have occurred due to usage by other systems prior to a demand by a system required for decay heat removal.

[

l -

RESPONSE

No other systems outside of the PSSS use water f rom the EWST during normal operation. During shutdown conditions, water is transferred from the EWST to the refueling canal. The design of the EWST is such, that during normal operation, that sufficient excess capacity O is provided to allow for water holdup in containment that could accumulate via operation of containment spray and still have l

O 440-5 AMENDMENT 1 l WAPWR-PSSS 7855e:1d MAY, 1986 l

sufficient NPSH for operation of the pumps using the EWST as a source of supply. Less than of the order of half of the capacity is held up l

prior to water returning to the EWST under those conditions.

1 1

440.34 One of the risk contributors encountered in some plant PRA reviews i 5.4.7 involves loss of pump cooling, such as is the case when the RHR pumps  ;

are cooled by component cooling water (CCW). What makes this event j particularly severe is that loss of CCW can also disable alternate cooling schemes as well if they require CCW for cooling. In support of CP Rule (1)(1) and NUREG-1070 guidance, please address this issue j with respect to the RHR - SI configuration in the SP/90. Of particular interest is possible use of the pumped fluid for component cooling, the consideration given by W to such alternate schemes, and the reasoning leading to the decisions taken with the SP/90 components.

RESPONSE

Motors used in the PSSS are air cooled using ambient air f rom the pump compartment. That air is, in turn, temperature controlled via the HVAC systems. Pumps are generally CCW cooled to cover the possibility that the pumped fluid could be hot under some accident conditions.

l 440.35 (p 5 . 4 -1, second paragraph) This indicates cooling to 150'F as  !

5.4.7 contrasted to 125'F in some of the prior submittals concerning W plants. What is the reason for the change in the end temperature l when referencing plant cooldown time? Please provide justification.

RESPONSE

9:

The selection of 150*F represents a more practical approach to plant cooldown. It permits more flexibility in design and operation, and also provides for potential cost benefits. W does not consider this to be a safety consideration or to have safety implications.

O WAPWR-PSSS 440-6 AMENDMENT 1 2855e:ld MAY,1986

O 440.36 What is the cooldown time from 350*F to 212*F in the SP/90 and how V 5.4.7 does this compare to other typical plants?

RESPONSE

Normal RHR cooldown is initiated at 350*F and is assumed to be started at 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> af ter nuclear shutdown. Initially the RHR flow through the RHR heat exchangers must be throttled to limit the cooldown rate of the primary system to 50*F/hr. Therefore there will be little difference, in the time to cooldown to 212*F between the SP/90 and a typical plant. For example; the expected cooldown for the SP/90 from 350*F to 202*F would take 4.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (8.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after shutdown), while the design basis cooldown for the Vogtle plant (3425 MWt) from 350*F to 204*F would take 3.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> (7.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> after shutdown).

440.37 What are expected CCW temperatures during the cooldown process in the

! 5.4.7 SP/90 plant and how do these compare to other typical plants?

O RESPONSE:

The expected CCW temperatures during cooldown for the SP/90 range from ~98'F at RHR initiation to ~87'F at the end of cooldown.

This compares to ~108'F and ~97'F for the Vogtle plant. This difference in CCW temperature is due primarily to the assumed essential service water temperature available during normal cooldown, which is 80'F for the SP/90 and ~91'F for Vogtle.

440.38 The third paragraph of p 5.4-1 begins "The low head pumps have

] multiple uses" and continues to list the uses as:

d 5.4.7

- conventional RHR pumps (cooldown, during refueling) l -

containment spray pumps long term ECCS

- water transfer during refueling (from the refueling canal and the Emergency Water Storage Tank (EWST)

O WAPWR-PSSS 440-7 AMENDMENT 1 2855e:ld MAY,1986 I - -. _ _ - _ _ . _ _ _ _ - _ _ _ _ _ _ - - . _ __

Please address the following with respect to these multiple uses (Review requirements pertinent to this request are identified in the SRP, system operation as pertinent to the CP rule, plant experience as identified in NUREG-1070, and operation of the plant under severe accident conditions as covered in NUREG-1070.):

a. Provision of suitable safeguards so that the RHR function is maintained during use (of some) of the RHR subsystem (s) for other duty.

b. Potential for loss of all containment cooling and primary system cooling due to loss of RHR pumps. This question is based on recent plant experience which has identified a potential problem regarding loss of shutdown cooling during RCS maintenance. A number of occurrences have resulted in which the RCS has been j partially drained, improper reactor coolant system level control has resulted, there has been an unplanned loss of reactor coolant inventc,ry, and the RHR pumps have been operated with an inadequate NPSH with resultant air or vapor binding and subsequent loss of shutdown cooling. In the discussion of this i potential problem, please provide the following types of 1

information:

l

1. The design or anticipated procedural provisions to maintain adequate RCS inventory, level control, and NPSH during all l operations in which RHR cooling is required 1

2. The provisions for rapid restoration of the RHR system to service in the event the RHR pumps become air or vapor bound l

3. The provisions to provide alternate methods of shutdown cooling in the event of loss of RHR cooling during shutdown maintenance. This should include coverage of situations such as when the RCS has been partially drained for steam generator inspection and maintenance.

O WAPWR-PSSS 440-8 AMENDMENT 1 2855e:ld MAY, 1986

c. Assessment of the need for interlocks in addition to those discussed in this Module, and other measures to assure satisfactory operation of each of the RHR subsystems when different simultaneous duties are required, as may occur, for example, during accident conditions, testing while shut down, or during shutdown and startup associated with refueling operations. (An example is need for simultaneous ECCS and containment spray.)

d. In regard to EWST level for the various operations and configurations and NPSH considerations, please provide the actual NPSH requirements of the pumps contrasted to what is provided and the safety margin which exists as a function of plant conditions. (Table 6.3-2 shows the NPSH requirement and the amount available for the horizontal pump. The vertical pump is not mentioned, nor are the conditions defined which are applicable to the given available NPSH. The discussion should also address the uncertainty associated with the NPSH requirement and should address the adequacy of the stated margin.)

e. Each of the four RHR subsystems is described as identical. What consideration was given to diversity of manufacturer or of design with the objective of decreasing the probability of common cause j failure and why was this approach rejected?

RESPONSE

a. The RHR system has no safety function with respect to Appendix K. Alignment is normally for the spray function, with sufficient flexibility for other applications. Use of the RHR system for ECCS is beyond the design basis, with a number of multiple failures required before the RHR would be the only cooling or injection system remaining.

b. With respect to containment, there are four safety related fan coolers which are designed for this function. Further, there are WAPWR-PSSS 440-9 AMENDMENT 1 2855e:1d MAY, 1986

a number of lineups which will cool the RCS, which in turn effectively is providing the cooling required for containnent to remain in a safe condition. With respect to the difficulties in existing plants, each of the RHR subsystems is provided with a complete interlock system, power is to be removed as part of the operating procedures under appropriate conditions, and multiple valves (with suitable interlocks) are provided. If, for some reason, an RHR subsystem were to become vapor bound, one can open test lines and drain lines, which should ' allow rapid restoration of the function. Further, there are four subsystems, and the loss of one is of little significance. The SP/90 also provides a number of combinations for backup to the RHR as well. The charging system, four SI pumps, accumulators, and the CRT's all can be used as backup sources of cooling water. An additional feature is the direct vessel injection configuration for the SI pumps and the CRT's.

c. Question withdrawn by NRC.

d. The NPSH available is sufficient when compared to the NPSH required. NRC should perform the evaluation on the basis of the plant that is described in the SAR, as opposed to previously described options. Any changes in the SP/90 design will be described at a later date in formal documentation.

e. No consideration was given to diversity of manuf acturer or of design for this portion of the system. The approach was to select hardware where there exists extensive experience and established reliability.

440.39 With respect to CP Rule (1)(i) and the guidance provided in NUREG-

! 5.4.7 1070, was consideration given to a separate containment spray pump which could serve in a backup mode in the event of an accident? (An j example would be a diesel driven pump which was fully independent of AC power.)

l WAPWR-PSSS 440-10 AMENDMENT 1 2855e:1d MAY, 1986 l

1

RESPONSE

O This was not considered.

most probable events, The primary approach was to provide for the such as LOCA, and to provide sufficient equipment and flexibility that the less probable events are covered. J Containment spray actually needed is a very unlikely event, and if equipment were provided solely for the purpose of spraying containment, it probably never would be used. This introduces a potential reliability problem. The preferred approach, which has i been followed in the SP/90, is to use an item of equipment that is subjected to normal operation and is not needed under the conditions of required use being considered here. An example is the RHR pumps, which are not needed under full power operation conditions, and are F

made available (and configured) for containment spray when in a power operation mode. This approach has the benefit of allowing one to economically put the investment into equipment that is potentially

) needed on a probabilistic basis, and to not provide additional l

investment for equipment that is unlikely to ever be needed, while at I the same time making sure the need would be met if it were to occur via application of equipment that is normally used for some other i task.

i 440.40 (p 5.4-1, last line) The reference to Section 6.3.2.2.7 is incorrect.

5.4.7

RESPONSE

The correct cross-reference is Subsection 6.3.2.2.6. Ar. amended page 5.4-1 is provided as Attachment 440.40.

440.41 (5.4-2, third paragraph) This paragraph contains the statements:

5.4.7 "The heat load handled by the RHRS during the cooldown transient includes residual decay heat f rom the core, RCS sensible heat, and 3

reactor coolant pump heat. The design heat load is based on the decay heat fraction that exists at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following reactor

O 2

WAPWR-PSSS 440-11 AMENDMENT 1 I

2855e:ld MAY , 1986

l shutdown f rom an extended run at full power." We have the following questions relative to these statements

a. Is the heat generation rate f rom an extended run at full p.ter the same as from a run for an infinite time at full power?

b. What method was used in calculation of the decay heat generation rate and does it include all contributors to energy production such as the actinidet as identified in 10CFR5d App. K, I.A.37

c. Is the reactor coolant pump (RCP) heat that which is generated from operation of all four RCPs?

d. In computing the RCS cooldown characteristics, is an actual decay heat load used, or is the " constant" design value of 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> used?

e. What is the meaning of the design heat load statement with respect to the decay heat rate at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> and where is it and where is it not used?

f. Please provide a summary of the calculations used in determining the characteristics.

RESPONSE

a & b) The most recent decay heat basis used in determining the SP/90 cooldown is based on ANSI /ANS-5.1-1979, Section 3.6.

This " simplified method for determinnig decay heat power..."

considers an infinite operating period and inherently includes contribution from actinides in its conservatism.

This basis may be modified in the future, to consider a specific 3-region core burnup when detailed evaluation of the core equilibrium cycle is completed.

O WAPWR-PSSS 440-12 AMENDMENT 1 2855e:ld MAY, 1986

c) Only one reactor coolant pump is run during RHR operation s until the RCS temperature is reduced to 160*F.

d) The actual decay heat versus time is used in determining plant cooldown performance.

e) The meaning of the statement that the RHR heat exchanger is based on the decay heat at 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> is that: this condition s typically represents the limiting condition for sizing the RHR heat exchanger since the AT between the RCS fluid and the CCW is relatively small. During cooldown and post-accident operation, the heat removal performance of the RHR heat exchanger is calculated based on the actual temperatures of the primary fluid and cooling water calculated to exist.

f) The RHR cooldown characteristics are determined using a verified computer code which considers the time dependent O decay heat input; other heat loads such as RCS sensible heat, RCP heat, and auxiliary CCW heat loads; and pertinent flows and equipment sizes for the RHR (portion of the ISS), CCWS and ESWS. This code models the interaction between the ISS, CCWS, and ESWS and determines the resultant RCS temperature vs. time.

440.42 (5.4-2, fourth paragraph) According to this paragraph, the RHRS is 5.4.7 designed to address the functional requirements proposed by Regulatory Guide (RG) 1.139, and the statement is made that "The cold l shutdown design enables the nuclear steam supply system to be taken from hot standby to cold shutdown conditions using only safety-related systems, with or without offsite power, and with the most limiting single failure within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />." Since this regulatory guide was issued for comment, and is not approved, conformance is clearly not a requirement nor is the guidance approved. Neverthe-less, since W has asserted that the plant conforms to the issued for ifAPWR-PSSS 440-13 AMEN 0 MENT 1 2855e:ld MAY , 1986

.-g- ,--------,,,----,,,,,-,-,-r- -

comment guidance, the staff wishes to understand the degree to which the proposed standard plant conforms, and the extent, if any, that conformance of the design to the proposed RG 1.139 guidance may have introduced competing (adverse) risks in the design.

In regard to this application of the issued for comment RG 1.139, we have the following comments and questions: ,

a. The applicable statement from RG 1.139 is "The systems should be capable of bringing the reactor to a cold-shutdown condition within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> following shutdown with only offsite power or onsite power available, assuming the most limiting single failure." Please explain how this is the same thing as taking the SP/90 f rom hot standby to cold shutdown, and show how the dif f erent wording is in compliance with RG 1.139 in light of the meaning of " hot standby" as defined in the standard Tech. Specs.

b. Provide a list of each of the criteria contained in RG 1.139 and show precisely how the SP/90 design compares to each of the criteria.

c. What is considered to be the most limiting single failure with respect to the 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> criterion and how was this determined?

RESPONSE

a. The meaning is that the systems will be capable of accomplishing the goal f rom a hot zero power (critical, operating temperature and pressure) condition.

b. The SP/90 meets the functional requirements of Reg. Guide 1.139 for " Safety grade cold shutdown as follows:

O

1) FUNCTIONAL i

l a & b) The systems / components required to accomplish cooldown are all safety grade, redundant, and can WAPWR-PSSS 440-14 AMENDMENT 1 5855e:ld MAY, 1986

l operate with only offsite power or onsite power i

O available.

With consideration of the worst single failure; i.e.

loss of offsite power with failure of 1 of 2 emergency diesels and subsequent loss of 2 of 4, 120V, AC and DC protection trains: the minimum equipment available and utilized are summarized below.

Support Systems 1 of 2 emergency diesels and its associated lE, AC electrical power train 2 of 4, lE,120 volt, AC and DC instrumentation and control trains.

2 of 4 CCW pumps, HX's, and their associated piping and valves 2 of 4 ESW pumps, and their associated piping and valves 8 oration and RCS Inventory Control 1 of 3 RCS pressurizer PORV's and its associated I

block valve provide RCS depressurization i

capability (to 1600 psig).

l l

' - 2 of 4 ISS HHSI pumps and their associated piping and valves from EWST'to the RV

- 1 of 2 ISS emergency letdown lines f rom RCS to l

l EWST, includes 2 solenoid operated isol. valves O

WAPWR-PSSS 440-15 AMENDMENT 1 2855e:ld MAY,1986

1 steam isolation valve per SG (each with 2 independent trip devices) isolate steam lines.

1 SG safety valve per SG available for main-taining RCS at hot shutdown condition.

Cooldown to 350*F via SG's

- 1 of 2 TD EFW pumps and 1 of 2 MD EFW pumps and associated piping and valves, provide SG makeup.

2 of 2 EFWST's provide SG makeup water

- 2 of 4 SG PORV's and block valves, provide controlled cooldown capability with associated i makeup.

1 of 2 SG overflow isolation valves and associa-ted piping on 2 SG's used to depressurize and remove sensible heat from "non-cooling" SG's.

Cooldown f rom 350*F to " cold" conditions via ISS RHR function 4 of 4 accumulator discharge line isolation valves closed or 5 of 5 accumulator N 2 vent valves opened to prevent accumulator delivery.

1 of 3 RCS pressurizer PORV's and its associated block valve provide RCS depressurization capability (to 400 psig)

- 2 of 4 RHR pumps, HX's and associated piping and valves, includes unisolating 2 RCS/RHR isolation valves in each of the pump suction lines.

WAPWR-PSSS 440-16 AMENDMENT 1 2855e:1d MAY,1986

i l

- 4 of 4 CRT discharge line isolation valves closed or 5 of 5 CRT N 2 vent valves opened to prevent CRT delivery.

NOTE: No actions outside of the MCR are required to accomplish the above actions /actuations, c) The RCS can be brought to " cold" conditions ($ 200*F) within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. This time includes the following i

considerations.

~2 hours, operator assessment

~E hours, boration of the RCS

- ~6 hours, initial cooldown to 350*F via SG's

- ~22 hours, RHR cooldown to 200*F

! Items 2, 3, 4, 5, and 6 of Reg. Guide 1.139 have been addressed in Module 1 or by the above.

c. Loss of offsite power, coincident with failure of one diesel, is considered to be the most limiting single failure since this removes the maximum amount of critical equipment from operation.

Note that some of the problems associated with the control of the release of water onto the containment floor, due to water addition and removal for boration, are alleviated in the SP/90 because of the EWST and the lines connecting to it from the RCS.

O' 440.43 (5.4-2, last paragraph) This begins "The RHRS is designed to be iso-5.4.7 lated from the RCS whenever the RCS pressure exceeds the normal RHRS 1

cut in pressure." Previously. W stated the RHRS to be placed in 4 operation when the RCS temperature and pressure were approximately 350*F and 400 psig, respectively. What is the precise value that is considered to be the normal cut in pressure?

RESPONSE

O Question withdrawn by NRC.

WAPWR-PSSS 440-17 AMENDMENT 1 2855e:1d MAY, 1986

440.44 (5.4-2, last paragraph) The last line on the page is "its opening if 5.4.7 RCS pressure is greater than approximately ... and to". What is a typical plus and minus value that W would use that is consistent with the "approximately"? I

RESPONSE

l Question withdrawn by NRC. l l

440.45 (5.4-3, top of page) W, in discussing the motor-operated valves in l 5.4.7 the suction line, indicates that they automatically close if the RCS I pressure exceeds a specified value. Table 6.3-2 shows the design pressure of the low head pumps on the suction side and the design  ;

pressure of the tube side of the miniflow heat exchangers. Please discuss these pressures and expected equipment response. Include consideration of the time required for the valves to close upon receiving a close signal if the RCS pressure exceeds the specified auto-close pressure and the pressure rise one can encounter in the RCS. Include both calculational considerations, with a clear statement of the assumptions, and a comparison to what has been 1 experienced in practice, with discussion of applicability to the SP/90 due to differences in design. Also please address SP/90 response if the suction line valves fail to automatically close.

Please address the apparent change in philosophy in that, unlike prior W designs, credit appears to be taken for fast closing RHR isolation to provide some measure of RHR overpressure protection.

How is the approach described here consistent with the last paragraph of page 5.4-27

RESPONSE

These topics are considered to be of a generic nature, and the approach is the same as on existing W plants. With respect to the last question, the wording is not completely consistent, but a clarification is immediately provided.

O WAPWR-PSSS 440-18 AMENDMENT 1 2855e:ld MAY , 1986

440.46 (5.4-3, third paragraph) Does W consider the statement "... relieve 5.4.7 the possible backleakage ..." to be the same as " . . . relieve the maximum possible backleakage ...

" that has typically been used in past submittals?

O RESPONSE:

Relief valve 9020(A-D) has a relieving capacity of 25 GPM. The check valves providing the pressure boundary between the ISS and RCS are p verified to have backseated (to have low / unchanged backleakage), such V that RCS leakage Tech. Specs. and ASME Section XI requirements are met. Valve 9020 has suf ficient capacity to relieve the "possible" backleakage or " maximum possible" backleakage that would be deemed acceptable for continued plant operation.

440.47 In light of the plant experience identified in NUREG-1070 as an 5.4.7 element of consideration, what is the basis for determining the maximum possible backleakage?

-- RESPONSE:

The valves are provided for thermal expansion, not backleakage, although they would serve that purpose also. Provision for backleakage is not necessary. The pipe is specified as "2501" which means it is designed for 2450 psi. Note the tubes are designed for i

2250 psi.

440.48 With reference to CP Rules (2)(ii) and (2)(v), what indication is pro-i 5.4.7 vided to alert the operator if the RHR relief valves open and what is the anticipated operator action with respect to this event?

l RESPONSE:

No direct indication of RHR relief valves being open is provided to

, m the operator. Indicators such as temperature of the line if the RHR l or the SI is not operating will alert the operator to something being ,

wrong. Note that the isolation valves are tested prior to the plant being placed in operation.

O .

WAPWR-PSSS 440-19 AMENDMENT 1 2855e:ld MAY,1986 i

l 440.49 (5.4-4, top of page) Figures 6.3-1 and 6.3-2 show a number of the 5.4.7 components of the RHRS to be cooled by water from an external supply. Please discuss the interactions between the cooling water (typically CCW) valves and containment isolation signals.

RESPONSE

This topic will be covered in Module 13. With respect to CCW, no CCW that is required for safety will be isolated upon containment isolation.

440.50 (5.4-4, Section a. RHR Inner and Outer Isolation Valves ...) This O

5.4.7 section provides a listing of the interlocks provided which must be satisfied prior to opening the isolation valves on the RHR line from the RCS. With respect to past plant experience and NUREG-1070 guidance,

a. A number of other connections are also provided which can be affected when these valves are opened, including some lines of relatively large diameter leading to the EWST and CVCS. What measures are taken to assure that the valves shown as closed on the P&ID are indeed closed?

b. The P&ID shows valve 9018 (normally closed) as connecting to CVCS letdown. The associated P&ID for the CVCS system is to be in a module which we have not received. Will this show the connection downstream of valve 9018 to be a common header with other valves which could serve as a connection between the RHR subsystems and could lead to flow from one RHR subsystem to another?

RESPONSE

a. A number of test lines and connections are provided which will provide for recognition of incorrect positioning during the testing that will be performed prior to placing the plant in power operation. A complete interlock system, as discussed in the module, is provided which should prevent most of the valves O

WAPWR-PSSS 440-20 AMENDMENT 1 2855e:ld MAY, 1986

f rom being placed in an incorrect position. Finally, even if there were a problem with one cf the subsystems, there are three others which provided more than sufficient response to plant upset and accident conditions.

b. Two RHR subsystems will be connected to the CVCS via a line with a normally closed valve (9018A/B) and a check valve. Thus, no backflow from one RHR subsystem to the other will occur.

440.51 (5.4-5, second paragraph) "The wide-range RCS pressure interlock for 5.4.7 both the prevent open and the autoclosure features on the inner isolation valves is independent and diverse from that provided to the outer isolation valves." In light of the guidance provided in NUREG-1070 pertaining to plant experience, the anticipated work on PRA in response to CP Rule (1)(i), and recent plant experience with failure .to protect low pressure systems f rom the high pressure that exists in the primary system, please discuss the independence and diversity so that we may more fully understand if there are O' inter-dependencies.

RESPONSE

The wide-range RCS pressure interlocks on the inner isolation valves are indepenaent and diverse from the interlocks that are provided to the outer isolation valves. Diversity of the interlocks is provided by use of the set of wide range pressure transmitters for the inner isolation valves of a model different than the set of transmitters for the outer isolation valves.

440.52 (5.4-5, Section b. EWST Suction Isolation Valves ...) In regard to 5.4.7 possible precursors to severe accidents and CP Rules (1)(i) and

. . (2)(11), and to assist in providing the reviewer a better understanding of plant behavior, please address the following:

a. The following statement is made in this section: " Interlocks are provided for the EWST suction isolation valves to prevent their O

WAPWR-PSSS 440-21 AMENDMENT 1 2855e:ld MAY, 1986

opening unless one of the two corresponding RHR nonnal cooldown suction isolation valves is closed." What is the valve action (and other actions as well) associated with a LOCA while an RHR subsystem is in operation?

b. An additional statement made is "These interlocks ensure that the valve cannot be reopened by operator action after the initiation of a normal cooldown operation ...

Does this mean there are other, non-operator, actions which can cause the valve to be reopened?

c. In discussing the EWST isolation valves, the statement is made "The valves may be closed by operator action from the main control board at any time." Is this statement with respect to alignment for RHR operation or in general?

RESP 0NSE:

a. The P signal is blocked prior to reaching RHR entry conditions, and spray is blocked. Power lockouts are provided so that I

inadvertent valve operations are precluded. NRC postponed discussion of the portion of this item pertinent to operator actions until later in the review.

b. No.

c. In general.

440.53 (5.4-5, c. Containment Spray Header Isolation Valves ... ) This 5.4.7 section contains the statement " Interlocks are provided for the series containment spray header isolation valves to prevent their opening unless one of the two corresponding RHR normal cooldown suction isolation valves is closed. These interlocks ensure that the valves cannot be reopened by operator action after the initiation of a normal cooldown operation ..."

O WAPWR-PSSS 440-22 AMENDMENT 1 2855e:ld MAY, 1986 l

a. Are there non-operator actions that can cause the valves to be reopened? For example, what would happen if there were a signal due to containment overpressure that would actuate containment spray under normal RHR operation conditions?

O b. Section b. used the words "the main control board". This section uses "one main control board." Is there more than one? If so, which is being referenced here and why?

O RESPONSE:

a. See response to 440.52.

b. There is only one main control board.

440.54 (5.4-6, Section d. System Test Header Isolation Valves ...) The 5.4.7 statement is made "These interlocks ensure that the valves cannot be reopened by operator action after the initiation of a normal cooldown l operation..." Are there other non-operator actions that can cause the valves to open?

RESPONSE

See response to 440.52b.

440.55 (5.4-6, Section e. High Head Pump Discharge Header Isolation Valves 5.4.7 ...) " Interlocks are provided for the high head pump discharge header isolation valves to prevent their opening unless one of the two j corresponding RHR normal cooldown suction isolation valves is closed. These interlocks ensure that these valves which are normally open during power operation cannot be reopened by operation action after the initiation of a normal cooldown operation."

a. Are there other non-operator actions that can cause the valves to open? (We assume " operation" means " operator".)

O WAPWR-PSSS 440-23 AMENDMENT 1 2855e:1d MAY, 1986

b. What actions occur if there is an injection signal causing activation of the high head pumps while the RHR subsystems are in the normal operation mode?

c. "The valves may be closed by operator action from the main control board at any time." Is this statement specific to when operating in the RHR mode, or does it apply in general? For example, does it apply if operating in the SI mode with the high pressure pumps in operation and injecting water into the RCS?

RESPONSE

See prior questions and responses.

440.56 (5.4-6, f. High Head Pump Flow Control Valves ...) The above 5.4.7 questions for Section e. apply also to this section.

RESPONSE

( See prior questions and responses.

440.57 (5.4-7 Section 5.4.7.2.5 System Reliability Considerations) "The RHR 5.4.7 subsystem of the ISS is a four train, fully redundant safety-related system. ... Each train is physically separated and protected where necessary so that a single event cannot initiate a common failure."

Please discuss these statements with respect to:

a. Loss of one diesel in the two train diesel configuration with off-site power unavailable.

b. Potential flooding. Include consideration of the sumps shown in Figure 1.2-2 at the elevation of the rooms containing the RHR l pumps (Do they communicate and provide a potential flow path?)

and of flooding from above via the openings or stairways shown on the Figure. A potential path of concern for the latter is the openings and stairway paths shown on Figure 1.2-2 from the O

WAPWR-PSSS 440-24 AMEN 0 MENT 1 2855e:1d MAY, 1986

elevation above the RHR pumps, which appear to provide a communication path which circumvents the wall between the pump vaults and the valve rooms above. -

I

RESPONSE

a. Loss of one diesel with off-site power unavailable will remove 1 two of the ISSS subsystems from operation. Two will remain.

b. Double doors and equipment hatches will be provided for isolation. The intent with respect to the doors is to provide sufficient strength that they will remain intact under flooding conditions. For example, with respect to the path identified in the question, the doors are intended to remain intact when the pipe / valve compartment above that location is flooded to the ceiling.

440.58 A number of the interlocks discussed in Section 5.4.7 are dependent 5.4.7 upon pressure sensing instrumentation. With consideration given to CP Rule (2)(xix) as well as to normal operational conditions, please discuss errors in this instrumentation and the potential impact upon the RHR equipment and operation. Include extremes of temperature and pressure which could occur in containment under accident conditions and provide the values of temperature and pressure which have been considered.

RESPONSE

The design is to be O

V This is considered to be a generic topic.

consistent with existing W plant practice.

440.59 (5.4-8,second paragraph) "The RHR subsystem in conjunction with 5.4.7 other ISS subsystems, ..., and the secondary side safeguards system

' provide a safety related means for borating and bringing the plant to cold shutdown conditions following any accident not involving a LOCA in excess of normal charging capacity. As such, the design of these systems addresses the requirements outlined in Branch Technical O Position RSB 5-1."

WAPWR-PSSS 440-25 AMENDMENT 1 2855e:1d MAY, 1986

a. What equipment constitutes the secondary side safeguards system that is referenced here?

b. Our understanding of the W meaning of "LOCA" is any break which results in a leak rate in excess of normal charging capacity, which we have been told is a rate in excess of the capacity of the two centrifugal charging pumps. Is this correct?

c. Please provide a complete listing of the criteria outlined in RSB 5-1 and compare the SP/90 features to these criteria.

Specifically identify any criteria that are not fully and completely met in every respect and discuss why they are not met.

d. Please provide a discussion of the procedures (see also CP Rule (2)(11) used to take the plant f rom normal operating conditions to cold shutdown conditions which satisfy the following functional criteria:

(1) The design shall be such that the reactor can be taken from normal operating conditions to cold shutdown using only safety-grade systems. These systems shall satisfy General Design Criteria 1 through 5.

(2) The system (s) shall have suitable redundancy in components and features, and suitable interconnections, leak detection, and isolation capabilities to assure that for onsite electrical power system operation (assuming of f site power is not available) and for of fsite electrical power system operation (assuming onsite power is not available) the system function can be accomplished assuming a single failure.

(3) The system (s) shall be capable of being operated f rom the control room with either only onsite or only offsite power available. In demonstrating that the system can perform its O

440-26 AMENDMENT 1 WAPWR-PSSS 2855e:1d MAY, 1986

a ,_ -

function assuming a single failure, limited operator action outside of the control room would be considered acceptable if suitably justified.

, (4) The system (s) shall be capable of bringing the reactor to a cold shutdown condition, with only offsite or onsite power available, within a reasonable period of time following shutdown, assuming the worst limiting single failure.

4 O RESPONSE:

a. This is covered in Module 6/8.

b. This is essentially correct. It is not important when it comes to actual operation of the plant.

c. The criteria of RSB 5-1 are fully met by the systems / components

! listed in response to question 440.42 concerning compliance with Reg. Guide 1.139.

l I

j d. These topics have been covered in the response to prior questions.

440.60 (5.4-8, last paragraph and top of following page) "During plant 5.4.7 shutdown, the 'S' signal must be manually blocked by operator action

... to prevent an automatic ... actuation during shutdown ope rati ons . . . . When the RCS pressure has been decreased to approximately 1000 psig, the accumulator discharge isolation valves are closed by the operator ... and the valve breakers are O

administratively removed or opened to prevent an inadvertent opening 4 of these valves." Please discu s the operations required to restore operation of these portions of the SI system in the advent they are needed. (Reference CP Rule (2)(11).)

I I

RESPONSE

Restoration is merely the reverse of the procedures.

O WAPWR-PSSS 440-27 AMENDMENT 1 2855e:ld MAY, 1986 4

..n ,- -

,----+-.------,_--~,--r, -

-,--,.,.--a . , , - , - - - - - . . - - - , - , - - - - . - - - , - , . _ . - _ _

440.61 (5.4-9, second paragraph) "Following this (the above) the core 5.4.7 reflood tank isolation valves ... are closed by the operator f rom the main control board and the valve breakers are administratively removed or opened to prevent inadvertent opening of these valves."

Please consnent on the operations which are required to restore operation of the affected portion of the SI system in the advent it is needed. (Reference CP Rule (2)(ii).)

RESPONSE

See 440.60 response.

440.62 (5.4-9, fourth paragraph) " Initiation of the RHR operation includes 5.4.7 an initial warm-up period during which each RHR subsystem is aligned to circulate coolant from the low head pump, through the RHR heat exchanger through the RHR hot leg injection line, through the RHR inner and outer isolation valves and back to the low head pump suction."

a. Are these operations performed on each RHR subsystem in turn, with one being placed in operation prior to beginning warmup of the next, or are all RHR subsystems warmed up at the same time?

b. Does W consider that the guidance of RG 1.47 is applicable?

c. What is required to go to the SI mode if such an operation is necessary during this warmup operation? Spray mode? The intent of this request is not to solicit a list of operations. Instead, we are interested if any problems arise with respect to inter-locks or administrative operations and whether the necessary procedure will be covered in emergency operating procedures that W would recommend be used with the SP/90. These concerns are related to potential accident precursors for severe accidents as described in NUREG-1070 and CP Rule (2)(ii).

O WAPWR-PSSS 440-28 AMENDMENT 1 2855e:ld MAY, 1986

l

d. With respect to valve repositioning af ter the warmup period, we O note there is no longer any positive indication of the flow rate through the injection line, and that failure to completely achieve the necessary realignment could result in decreased or )

little flow through the injection line with no indication to the O operator of this condition with respect to the actual flow rate.

The original W submittal contained a drawing which showed a flow meter in the injection line. This would eliminate this potential problem. Why was the flow meter removed in this later drawing

, submittal? Does W plan to put it back in? Does the design, in W's opinion, meet the guidance of RG 1.97 without this capability? Note also CP Rule (2)(ii), (2)(v), and (2)(xix).

RESPONSE

i

a. Each subsystem can be warmed up sequentially.

b. Reg. Guide 1.47 excludes the need to provide bypass indication for inf requent but expected operations, like RHR cooldown. In O' fact, the automatic initiation of safety injection is bypassed throughout the plant cooldown. However, the ISS with four RHR i subsystems permits the operator to have greater flexibility during RHR cooldown to maximize the capability of the ISS to respond to unexpected situations. For example, during the early portions of RHR operation or when the temperature of the essential service water to the CCW HX's is " cooler" than design; one ISS subsystem can remain in its safety injection alignment so that SI and containment spray are available upon manual actuation of the S-signal, with no impact on the refueling critical path,

c. The only operations that are necessary are to close the inner and outer isolation valves. This removes the interlock restrictions, and SI and spray mode then can be used.

O WAPWR-PSSS 440-29 AMENDMENT 1 2855e:ld MAY, 1986 l

d. The meter was removed at the same time as a bypass line was modified. There are no plans to put the meter back into the design. W opinion is that the guidance of RG 1.97 is met without this capability.

440.63 ( 5.4-10) The second paragraph contains the statement "During this 5.4.7 warm-up period, the component cooling water system would not be aligned to deliver to the RHR heat exchangers or to the low head pump miniflow heat exchangers." The bottom of the page contains "(4)

Verify that component cooling water flow to the RHR heat exchanger has been established automatically on receipt of actuation signal initiated by temperature transmitter ... Please provide a brief discussion of transient stress considerations applied in designing the RHR heat exchangers and compare the mode of operation in the SP/90 to typical W plants of older designs. (See also 440.65.)

RESPONSE

Withdrawn by NRC.

O 440.64 (5.4-10, last paragraph) "To maintain high purification flow rates, 5.4.7 low pressure letdown is established from two of the RHR subsystems via the CVCS low pressure letdown isolation valves (9018A/B)." Only two valves are indicated with an "or" between. This does not appear consistent with the statement that two will be used. Please explain. In addition, if only subsystems A and B are to be used, what is to be the procedure if one or both of these subsystem trains is not in operation?

RESPONSE

O l

l This is really a plant availability issue. It is more appropriate to Module 13.

440.65 (5.4-11, first paragraph) "The operator can adjust the allowed cool-O l 5.4.7 down rate by manually regulating the flow to the RCS. This can be accomplished by reopening one or more of the RHR hot leg O

WAPWR-PSSS 440-30 AMENDMENT 1 2855e:1d MAY, 1986

i i

1 3

l recirculation header isolation valves ... and/or by closing one or l

more of the reactor vessel injection header isolation valves ....

4 These operations establish either a full or partial flow bypass mode where the RHR pump flow is redirected from the reactor vessel i injection path to the RHR pump suction line."

a. These operations appear to result in a cold RHR subsystem due to continued flow through the heat exchangers with CCW providing cooling. This would appear to defeat the purpose of the RHR subsystem warmup procedure unless the CCW is automatically j controlled to control the RHR subsystem temperature. Please

explain. If the RHR subsystem is cooled by this procedure, i

j include in the explanation a comparison to other W plants where the bypass flow bypassed the RHR heat exchanger rather than the f RCS.

f Please discuss temperature control since there is no direct f

b.

i indicator of flow rate in the injection line, as discussed in a i

prior connent and questions.  :

} '

c. Please discuss what subsystem control failures can cause cooldown at greater than permissible rates and assess the safety impact, if any, of such conditions.

d. Are the header isolation valves designed to permit flow control j or are they essentially on/off valves?

i

RESPONSE

a. The RHR system would not be operated in this mode, it would be I turned off instead.

O b. A direct indication of flow in the injection line is not necessary to provide proper control.

O 440-31 AMENDMENT 1 WAPWR-PSSS 2855e:ld MAY, 1986

c. There is no safety impact due to control system failures causing cooldown at greater than permissible rates.

d. These are presently on/off valves, although this may be changed.

440.66 (5.4-11, second paragraph) Prior W design descriptions include a 5.4.7 discussion of water solid operation during cooldown. Is this operational mode proposed for the SP/90? If so, at approximately what conditions during cooldown is it anticipated to occur?

O

RESPONSE

The intent was that the SP/90 not be operated in a water solid

condition.

440.67 (5.4-12, third paragraph) This paragraph describes how the RHR pumps 5.4.7 can be used to drain the refueling cavity to the top of the reactor vessel via flow f rom the suction lines into the EWST via the full flow test line isolation valves. However, Section 5.4.7.2.1 d.

describes an interlock to prevent opening of the test valves unless one of the suction line valves is closed. Moreover, Section 5.4.7.2.1 a. describes an interlock which prevents opening of the suction line valves unless the test valves are closed. This appears to block the planned flow path. Please explain.

RESPONSE

The interlock in question is provided with 'a special feature which permits the operator to open the full flow test line isolation valves during refueling cavity draindown. This feature consists of an administratively controlled and alarmed perTnissive switch which must be held in the "RHR to EWST TRANSFER" position while each test line isolation valve is opened.

O 440.60 (5.4-12, Section c. Plant Startup) "When plant startup begins, low 5.4.7 head pump operatirin would be terminated. However, one RHR subsystem would remain alitned to the RCS to maintain a low-pressure letdown O

WAPWR-PSSS 440-32 AMENDMENT 1 2855e:ld MAY, 1986

N i

path to the CVCS. This alignment provides RCS pressure control while the pressurizer heaters are forming the steam bubble and heating the pressur'izer. As the reactor coolant pumps are started, their thermal input begins heating the reactor coolant. Once the pressurizer steam f bubble formation is complete, the low-pressure letdown path to the CVCS is isolated by closing the letdown isolation valve.... At this point in the startup operation, all four RHR subsystems would be isolated." One of the difficulties with previous PWR designs has been overpressure under startup conditions with the RCS solid. There are many similarities between the SP/90 design and prior M designs in this respect.

l

a. Please discuss the means to be taken with the SP/90 as presently configured to avoid overpressurization under all conditions of temperature and pressure which allow the RHR system to be connected to the RCS, Include consideration of relief capacity, d

valve response times, and your selection of the most severe overpressure event with consideration of operating experience.

(See also 440.45.)

b. Did M consider leaving more than one RHR subsystem connected to the RCS for pressure control during these startup operations? If

.so, why was this operational mode rejected? If not, what are the reasons for not considering this mode?

i c. One of the items addressed in NUREG-1070 is the application of past experience to design of new plants. In light of this O guidance, what considerations were given to low temperature overpressure protection in the process of designing the SP/90 and what led to the existing configuration and operational technique?

I i

d. The above questions address the problem of overpressure. A i

closely related problem is RCS pressure undershoot following response to an overpressure condition and the potential effect upon RCP seals, which are designed for a specific range of WAPWR-PSSS 440-33 AMEN 0 MENT 1 2855e:1d MAY, 1986

operational conditions, and the low pressure end of the range may be exceeded. Please address the consideration given to this situation in the SP/90 design process.

We plan further review of this area when we review Section 5.2.2, -

which was not provided in Module 1. Insofar as the responses to the above are adequately covered in that Section, a reference to the Section is sufficient reply.

RESPONSE

This item was postponed until review of Section 5.2.2.

440.69 (5.4-13, third paragraph) "When the RCS pressure exceeds the core 5.4.7 reflood tank and accumulator pressure, the core reflood tank and accumulator discharge isolation valves ... are opened...." Does this mean the core reflood tank isolation valves are opened when the RCS pressure exceeds the core reflood tank pressure and then the accumulator discharge isolation valves are opened when the RCS pressure exceeds the accumulator pressure?

RESPONSE

Yes.

440.70 (5.4-13, last paragraph) In regard to normal operation and the 5.4.7 initiation of normal operation, the statement is made that "All ISS piping is completely filled with borated water." What steps are taken to assure this is the case?

RESPONSE

SP/90 plant operating procedures will assure that the piping is completely full.

O

! 440-34 AMENDMENT 1 MAPWR-PSSS MAY,1986 2855e:ld

I q

440.71 (5.4-15. Table 5.4.7-1, Failure Mode and Ef f ects Analysis - Integrated 5.4.7 Safeauard System ISS) Residual Heat Removal Normal Cooldown Function

- Active Failures) We note these are all failures of a single item to perform a function at the time it is called upon to initiate an I operation. There appears to be no consideration of failures during operation. In addition, there appears to be no consideration of common cause failures by which a single failure at some point outside i of the RHR system could cause more than one of the RHR subsystems to j fail. An example is failure of a diesel to start following a loss of off site power, which would appear to affect two RHR subsystems in the two diesel configuration.

a. Are we correct? If so, please address common cause failure modes.

t l

b. Please identify which of the RHR subsystems are connected to which of the diesels for the two diesel plant configuration. In I

particular, address which diesel (s) are connected to RHR subsystems A and B and, if both A and B are connected to the same

. diesel, discuss the implications with respect to A and B being the only RHR subsystems with connections to CVCS letdown, i

j RESPONSE:

a. See 440.181.

l

b. The RHR subsystems are powered by the diesels located directly f

i above them.

1 4

i 440.72 (5.4-15. Table 5.4.7-1) Please address the items identified below:

5.4.7

! a. Item 1 "Effect on System Operation," states "The valves are located outside containment and action taken to manually close

.O them will restore subsystem operability."

P&ID with respect to valves 8813 and 8814 and their operation and Please examine the discuss your findings.

O

! WAPWR-PSSS 440-35 AMENDMENT 1 2855e:ld MAY, 1986

+ , . --n --. . ww~- ,--,w,-,.. ,_.an-w- . - -

---v- .,r--.wn m v--,.-- , w,.,-,..,-- m n m,- ,- -- =,,w

b. Item 1 does not list the normally closed valve 9011. Does this mean that for this example, and in general for the entire table, you do not consider an active failure to occur if a valve is mispositioned?

l

c. Why is valve 9015 failure to open on demand not addressed?

d. Why is valve 8810 failure to open on demand not addressed?

e. Why is valve 8810 failure to close on demand not addressed?

f. Why are the effects of valves failing to open or close on demand not discussed with respect to cooldown rate control?

l I

g. Why is this table restricted to only the normal cooldown function? A number of other operational modes are of concern with respect to active failures in the RHR system, such as plant startup and overpressurization associated with solid plant operation.

h. Item 2, in the " Failure Detection Method" column, contains references to coolant flow indication via FI-920. This item, although it appeared in preliminary P& ids, is no longer contained in the P&ID submitted with this module. Please explain.

i. Item 3 also contains references to FI-920. See above item h.

1

j. Item 4, which addresses valve failure effect on letdown to the CVCS and the implications with respect to boron concentration, does not provide information as to whether there is any effect of RCP operation as opposed to their not being in use. Does this consideration have an impact on your conclusions?

k. In item 4, the " Failure Detection Method" column contains a l reference to flow indication via F I-132. Where is this indichtor located on the P&ID?

WAPWR-PSSS 440-36 AMENDHENT 1

(

2855e:ld MAY, 1986 l

l __ __ - __ _

1. In Item 5, the motor operated valve failing to open on demand is O- indicated as detectable via flow indicator FI-920, which as previously identified is not provided according to the information supplied to the staff. Note that if it were provided, there should be a reference to it in the failure to O close on demand entry, which appears to indicate an inconsistency within the table.

RESPONSE

O. a. Valves 8813A (B.C.D) and 8814A (8,C,0) are the full flow test line isolation valves and would normally be closed, and would not need to be moved to align the RHR flowpath. They should be deleted from Item 1 of Table 5.4.7-1. Also note that valves HCV-858 (859, 860, 861) have been deleted in subsequent versions of the ISS. If NRC deems this necessary this table will be updated for the integrated PDA submittal, w b. Yes. Mispositioning of a valve is not consideret to be an acti te s failure.

c. NRC will address the need for this table. If NRC decides the table is necessary, then H will address the item.

d. See item c, above.

e. See item c, above.

f. See item c, above.

g. See item c, above.

h. See item c, above. W will provide a correction if the table is retained.

i. See item h, above.

WAPWR-PSSS 440-37 AMENDMENT 1 2855e:ld MAY,1986

j. This topic is addressed in Module 13.

k. This information will be provided in the CVCS P&ID.

1. As previously discussed, the flow indicator has been removed from the design. The other indications to the flow meter in the table should not appear.

440.73 Figure 6.3-1. ISS Piping and Instrumentation Diagram, Sheet 1, shows 5.4.7 several pressure relief valves. What indications are to be provided to the operator regarding their status?

RESPONSE

No direct indications will be provided to the operator. Note THI item II.D.3 applies only to the RCS.

440.74 Figure 6.3-1 in part contains the RHR system P&ID. We have the 5.4.7 following comments and questions with respect to sheet 1 of the figure:

a. Would a break in the RHR pump suction line cause water to drain f rom the EWST into an uncontrolled region outside of containment when the containment is at atmospheric pressure? If yes, and the volume into which the water were to be draining were to fill, would the leak then stop? What would be the boundary of the volume under those conditions? Please also address these questions for a condition of elevated pressure in the containment. A portion of the concern addressed by this item is upcoming consideration of severe accident response, precursors to accidents, and decontamination factors that may be applicable for certain accidents involving leakage or bypass of containment.

O

b. Connections to piping in other drawings are indicated by reference to a drawing number. No drawing numbers are provided on the drawings. The drawings also do not have figure or sheet O

WAPWR-PSSS 440-38 AMENDMENT 1 2855e:ld MAY, 1986

l 1

)

m numbers and are identified in the written material as preliminary. Please correct and resubmit this material. Any changes between what we have reviewed in these preliminary drawings and in the final versions should be clearly identified.

c. Drawing legibility is poor, and in portions we have to guess at detail such as valve numbers: This makes review dif ficult.

Please resubmit these drawings in a legible form.

O d. What actions occur with respect to the relief line downstream of valve 9022 upon containment isolation?

e. Valves 9022 and 8850 (we could not read these valve numbers from the drawing, and obtained them instead f rom Table 6.3-3) have different opening pressures and protect components with different design pressures. The stated backpressure under normal conditions and the relief capacity are given. We have the C) severe accident conditions:

(1) What is the backpressure if the containment is at atmospheric pressure and both valves are open?

(2) What is the behavior with respect to opening and backpressure if containment is at 125 psig?

(3) What is the reasoning between selection of the relief pressures that in one case are at the design pressure and in the other are below design pressure?

l (4) What are the tolerances with respect to opening pressure for these valves?

l l (5) What criteria were used and what is the background for the criteria with respect to selection of the relief rates?

l WAPWR-PSSS 440-39 AMENDMENT 1 2855e:ld MAY, 1986 1

l

f. In the nonnal RHR operating mode or the low pressure injection mode, the flow rate indicated to the operator is the pump gross flow rate. Included within this flow rate is the miniflow from the pump outlet to the pump suction side. In the case of the high head SI pump, separate flow indication is provided for the miniflow rate and for the remainder of the pump output. Why are the two systems treated differently in that there is no miniflow indication provided for the RHR pump? Note the thrust of the question is toward providing a true flow rate for operation when near the shutoff head of the RHR pump as opposed to there being a control valve for the miniflow for the SI pump without one for the RHR pump. Knowledge of these flow rates may be important when addressing topics covered by CP Rules (2)(11) and (2)(xix).

g. Are the flow instruments that provide the RHR flow indication as discussed in the above item f simply flow indications to the operator or are flow rates provided?

h. Valves 9000 and 9001 are located in series and valve 9000 is in parallel with check valve 9019 which provides protection from overpressure between valves 9000 and 9001. Considering the l

NUREG-1070 instructions to consider past experience:

(1) Did W consider providing pressure indication or some other means of determining valve leakage for the region between valves 9000 and 90017 (Note such means could perhaps also serve as a backup indication for valve position.) If not, j

why not? If so, why was the provision rejected?

l (2) Why are test headers and connections not provided for the line connecting valves 9000 and 900l?

O

1. Please discuss containment cooling for the case of operation in the containment spray mode. In particular, discuss the means of cooling the spray water f rom the time it leaves the tank until it is ejected into the containment.

WAPWR-PSSS 440-40 AMENDMENT 1 2855e:ld MAY, 1986

j. What is the design pressure of the line between valves 8810 and 8811? The concern is how pressure is relieved when the only relief path is against RCS operating conditions.

k. If valve 8810 is open as opposed to the normally closed position, flow is bypassed when the RHR pump is in operation, and SI injection flow is not as generally planned. What potential problems would this cause in expected operation of the RHR and 51 systems, are they considered to be of concern, and if so, how are they to be prevented? Please consider both the normal licensing concerns with respect to this item and the broader concern as discussed in NUREG-1070.

1. We are having difficulty identifying the connections to the test header, test line, and reactor vessel insofar as locating them on the referenced drawings is concerned. This may be due to having to guess at the nomenclature due to the poor quality of the drawing. Please provide the information in a legible form and O' indicate more specifically which drawings are referenced since the sheet number is not provided on the drawings.

m. Note 3 reads "See system standard design criteria 1.14, containment isolation in reference (illegible)." What is this and where may we find it?

RESPONSE

a. When the containment is at atmospheric pressure the level that O~ compares to the level in the EWST is roughly two to three feet up the door in the pump compartment. As previously discussed, the doors are designed to withstand a head of water. For a condition of elevated pressure in containment, it is not designed to O protect against the type of damage that may occur if water is released through a broken pipe at high pressure.

b. Revised drawings with system interconnections will be resubmitted.

O 440-41 AMENDMENT 1 WAPWR-PSSS 2855e:1d MAY, 1986

c. Revised drawings with system interconnections will be resubmitted.

d. None.

e.(1) Valve response is a function of the differential pressure across the valve. Hence, if there is a significant backpressure, the opening pressure of the valve will be affected accordingly.

e.(2) See above.

e.(3) Deleted by NRC.

e.(4) W considers this to be a generic item. Standard components are used in the SP/90 of the type used in existing W provided plants. They are regulated by the codes, tested, and periodically retested.

e.(5) Almost all were selected in accord with existing practice.

They are generally provided for thermal expansion caused i

pressure release.

l l

f. The pumps are selected to provide a particular flow rate at a selected pressure, and flow rates above that pressure are not of interest with respect to normal plant operation. If one wanted the pump to deliver above that pressure, then the pump selection would have been made so that that would be the case. Hence, actual flow rate in the vicinity of the shutoff head with miniflow is not of interest provided the design criteria are met. In the case of the RHR pump, the normal usage of the pump if for RHR duty, and the assigned duty when the plant is at power is for use as a containment spray pump: Use in an SI mode is a very unlikely backup. With respect to the differences in piping, the RHR pump miniflow always takes place if the pump is in operation, and the flow merely circulates from the discharge side MAPWR-PSSS 440-42 AMENDMENT 1 2855e:Id MAY, 1986

of the pump to the suction side through a heat exchanger. In the s case of the high head SI pump, the miniflow is back to the EWST, thus providing a path whereby flow does not remain within the SI pipeline, and there is also a valve in the miniflow line.

4

g. Flow rates are provided.

h.(1) These were not considered and they are not required.

h.(2) W will consider installing these headers and connections.

i. Such cooling is not needed. Note the containment coolers are safety related equipment and will perform the containment cooling function. In addition, if for some reason, cooling of the spray water was needed, then one or more of the RHR subsystems can be operated in a lineup which provides for flow through the RHR heat exchangers, and the others can be operated in the containment p spray mode.

V

j. This is designated as a 2501 line and therefore will withstand RCS pressure.

k. The valve positions will be indicated to the operator in the f

advanced control room display. Even if the valves are in the wrong position in one of the RHR subsystems, it doesn't hurt anything because the other three subsystems are independent and will function normally.

1. Revised drawings with system interconnections will be resubmitted.

m. This is a standard note which refers the user to the flow diagram legend sheet and to the W System Standard Design Criteria 1.14, which provides additional arrangement detail for containment isolation. This criteria and others are provided to the customer as part of a " Standard Information Package" (SIP) and has no l O specific application to the SAR.

440-43 AMENDMENT 1 MAPWR-PSSS 2855e:ld MAY,1986 l

440.75 Are there any systems or components needed for shutdown cooling which 5.4.7 are de-energized or have power locked out during plant operation and which have not been identified and discussed in Section 5.4.7? If so, what actions are going to be recommended to restore operability when they are needed? A portion of the concern is pertinent to severe accident conditions, operations for off-power conditions, and CP Rule (2)(11).

RESPONSE

There are no such systems or components.

440.76 Are there any concerns with respect to freezing involving any of the 5.4.7 systems or components described or mentioned in Section 5.4.77 If so, please identify them and provide a discussion.

RESPONSE

There is no concern with freezing. All components are located within containment or housed within a structure that is controlled by the HVAC systems.

440.77 Please describe the actions which take place if the RHR pumps are 5.4.7 running in response to a LOCA condition in the RCS, and the contain-ment pressure becomes elevated so that containment spray is automati-cally required. This information is considered applicable to CP Rule (2)(11) and some of the of f-normal conditions of operation that may be encountered under some of the situations covered by NUREG-1070.

RESPONSE

The staff reviewer is correct in referencing this as an of f-normal condition. Under normal circumstances, the RHR pumps are used only for containment spray or' in the RHR mode, and their use under an SI configuration would be very unusual. However, if the postulated situation were to develop, the spray function would be initiated by the P signal causing the spray valves located on the discharge side of the RHR pumps to open. This would provide a flow path to the i

MAPWR-PSSS 440-44 AMENDMENT 1 2855e:ld MAY, 1986 i

spray headers in addition to the path via the S1 lines to the RCS.

Some flow would be expected to go to each of the paths. There is a possibility that pump runout would be a potential problem.

440.78 Please describe the consideration given to loss of water from the EWST v 5.4.7 and the conclusions with respect to the likelihood of this occurrence in the SP/90 design. (CP Rule (1)(i) and potential with respect to severe accident situations as covered in NUREG-1070.)

O

RESPONSE

This is considered to be very unlikely. If it were to occur, two general types of paths would be followed. One would be a leak into the pump compartments, and the other a path via a small line such as a flow purification line. The latter, of necessity, would have to be small since it would be restricted by line size. Further, the line elevation is such that the line goes above the level of the EWST.

The first path would provide leakage into one of the pump compartments. If small, this leakage would be pumped out of the O' compartment by the sump pumps and the operators would be alerted to a problem. If large and for some reason it could not be controlled, the leak would fill the compartment, as discussed in the response to 440.74a, and then would stop. Also note the EWST is provided with multiple level measurement devices, and alarms are sounded if the level does not meet requirements.

440.79 Was consideration given to a means for reducing RCS pressure to signi-5.4.7 ficantly below the RHR initiation point (such as a large blowdown O valve)? Please discuss in light of the guidance to consider past experience that is provided in NUREG-1070.

RESPONSE

O, Yes. Three PORV's and two blowdown connections are provided which essentially lead from the RCS to the EWST. Connections are also provided which lead from the steam generator secondary side to the i

O WAPWR-PSSS 440-45 AMENDMENT 1 2855e:ld MAY, 1986

EWST which could be useful for this purpose, although their primary function is to provide a backup means for control of steam generator inventory under steam generator tube rupture conditions.

440.80 A recent paper on improvement of light water reactors in Japan con-5.4.7 tained the statement " Operability will be improved by separation of systems and equipment by function through such measures as discon-tinuing the conunon use of the ECCS for the containment cooling system and the shutdown cooling system. This will also contribute to improvement of reliability." Please comment with respect to the SP/90 design, CP Rule (1)(i), and the reasons for going in the selected direction as opposed to the above recommendation.

RESPONSE

See the prior discussion in regard to selection of equipment. W does not agree that the equipment should not have multiple uses and that setting up equipment for only one usage leads to better reliability.

In the case of the ISSS, the SI pumps are assigned to essentially that duty. The containment spray function is considered to be a far less likely event, and use of the RHR pumps for this purpose (the only assignment for these pumps when the reactor is at power) is considered to be sufficient. Note also that few operations are required for the safety functions to be met, as contrasted to existing plants, where more operations are required for satisfactory long term operation of the systems.

440.81 We note that all of the large lines penetrating containment have 5.4.7 valves that are remotely operated that are located outside of containment... except for the four RHR suction lines. Was consideration given to providing such a valve in each of the suction lines as a means of isolation which would provide backup for isolation? Please discuss with respect to past experience as outlined in NUREG-1070 and the reasons for the approach followed in the SP/90 design.

MAPWR-PSSS 440-46 AMENDMENT 1 2855e:ld MAY, 1986

i

RESPONSE

Consideration was given to providing such a valve. It was rejected because of the cost.

440.82 (6. 3-1, second paragraph) The listed functions of the ISS following 6.3 an accident include "... to provide negative reactivity (boroh) to bring the reactor subcritical; ... to remove the stored and fission product decay heat ...."

.O a. One of the listed functions is to bring the reactor subcritical.

But there is no statement in regard to maintaining the reactor

]

subcritical. Is this achieved?

b. In response to the statement in NUREG-1070 that "

... the applicant for certification of a reference de:;ign shall consider a range of alternatives and combinations of alternatives to address the unresolved and generic safety issues and to search for cost-effective reductions in the risk from severe accidents.", what consideration was given to alternate systems to provide emergency boration during the SP/90 design process, what such systems were considered, and what was the resolution with respect to each of the systems?

RESPONSE

a. Yes. It is covered in Module 6/8.

b. Existing plants have a large margin and there is significant l

O over-conservatism, particularly with the BIT tanks, which have l

been removed in some plants and have been a source of operational trouble in many of the plants which have them. No alternate

systems were considered for the SP/90 in addition to the systems which have been provided. The SI pumps are considered to be completely satisf actory for meeting the safety related equipment requirements for boron injection when coupled with all of the I other safety related equipment that is available for RCS depressurization.

l WAPWR-PSSS 440-47 AMENDMENT 1 2855e:ld MAY, 1986

440.83 Staff evaluations of the influence of vent valves in the reactor 6.3 vessel, such are installed in B&W plants, show significant advantages during reflood following LOCA as well as at selected other conditions following initiation of a small LOCA. What consideration was given to vent valve designs in the SP/90 and what was the process which led to their not being included in the design?

RESPONSE

Reactor vessel vent valves communicating between the upper vessel and the downcomer annulus were briefly considered in the SP/90 design.

They were rejected because of the cost, extra complexity, and the perceived difficulty of maintaining verification that they were in the correct position and would work when they were supposed to. The SP/90 has considerably more margin with respect to LOCA than is the case with present generation plants (large break LOCA temperatures are lower and the core does not uncover for small break LOCAs) and vent valves are not beneficial.

440.84 (6.3-2, last paragraph) The wording of this paragraph is identical to 6.3 to at least one SAR previously submitted for a typical W four loop plant, with modifications due to differences in the electrical and ECC train design and selections. However, the following wording is omitted from the SP/90 submittal: " Valves are provided in series where isolation is desired, and in parallel when flow paths are to be established for ECCS performance." Please identify each situation where a change in the above philosophy f rom prior plant design has been incorporated into the SP/90 design, if any, and provide the reason for the change. In addition, if applicable due to a change in philosophy, please discuss the safety implications (both positive and negative) of each situation identified above.

RESPONSE

With respect to the ISSS, no single line is used from a single set of pumps in the SP/90, and there are four SI subsystems. The need for l parallel valves has been eliminated. With respect to the charging 440-48 AMENDMENT 1

( WAPWR-PSSS 1 1 2855e:ld MAY, 1986 l l

l

system, plants such as the 412 plants have one discharge line from

' the pumps to the BIT tank, parallel valves, and then eventually a single line again. This was necessary to assure a reasonable probability of flow. The configuration does not exist in the SP/90.

(" With respect to series valves, they are used in the SP/90 in such a way that no single valve can be opened which will cause overpressuri-zation of a system leading to spillage. See also the discussion on the RHR system pertaining to the interlock systems.

440.85 (6.3-3. first paragraph) "All valves required to be actuated during 6.3 ECCS operation are either located to prevent vulnerability to flooding or are evaluated to ensure the consequences are acceptable.

The multi-train design of the ECCS in conjunction with power removal features for certain valves provides protection against repositioning of valves due to spurious actuation coincident with an accident."

t

a. Which valves are located so they could be exposed to flooding?

l t

b. Which valves have been evaluated with respect to the consequences of flooding and what were the results of the evaluation?

c. Which valves have power removed to provide protection against repositioning?

l

d. What are the safety impacts of power not being available to these valves in situations for which they may be needed?

RESPONSE

a. None for the systems that remain operative. A break of sufficient size in an ISSS subsystem to flood a compartment would j render the flooded subsystem inoperative, and the question of valve flooding response is moot. The important consideration is that the other three subsystems are unaf fected.

O WAPWR-PSSS 440-49 AMENDMENT 1 2855e:ld MAY, 1986

b. All of the valves inside containment have been evaluated. This process will be repeated again via study of the layout during the design process. The intent is to establish criteria to preclude this being a problem. The plan is that all valves will be above flood levels with the exception noted above in the response to part a of this question. Flood level is controlled by the weirs, and water that fills containment above the weir level spills back into the EWST.

c. It is not necessary to remove power from valves to provide protection against flooding.

d. Not applicable since power will not be removed. I 440.86 (6.3-3, third paragraph) "The elevated temperature of the emergency 6.3 water storage tank solution during recirculation is well within the design temperature of all ECCS components." What is the referenced elevated temperature?

RESPONSE

(a,c) The maximum peak calculated pressure is[]psig. The maximum temperature is assumed to be the saturation temperature that corresponds to that pressure.

) 440.87 (6.3-4, Section a) This Section contains a list of actions which are 6.3 initiated upon encountering an "S" signal. There is no mention of the charging system.

O

a. Are there any actions which occur with respect to any components in the charging system associated with the "S" signal? Please discuss, including the SP/90 plant behavior for a condition in which the RCS remains at a pressure above the injection pressure of the SI pumps, and for a condition where the pressure returns to a value above the shutoff head of the SI pumps.

O WAPWR-PSSS 440-50 AMENDMENT 1 2855e:1d MAY, 1986

b. There are a number of valves that are normally open, and which must be open for the SI system to function properly, that are not mentioned in the list. These include 8820A, B, C, D; 8803A, B, C, D; 8807A, B, C, 0; and perhaps HCV858, 859, 860, 861; and 8824A, B, C, and D. Does this mean there is no signal sent to these valves to open upon receipt of an "S" signal? If so, what is the reasoning which led to the decision not to signal these valves to open? Include how human error to properly align the system prior to possible demand has been considered. Section 5.4.7.2.6, Manual Actions, c. Plant StartuD, and Section 6.3.5.2.6, Monitor Lights. do not provide backup information to assure the steps are performed, do not provide assurance that an unanticipated valve operation is not performed at a later time, and do not provide detail as to which component positions and conditions will be displayed.

c. There are other valves that are normally closed where apparently no signal is sent to assure they are closed. An example is valves 9018 leading from the SI pump system discharge prior to entering the RCS and leading to the CVCS letdown. This is identified as a relatively large line and if one or more of these 1

valves were open, it would appear to defeat the function of the associated subsystem. Please discuss with respect to these valves and any others (such as 8810) where an incorrect position could have a serious affect on the SI subsystem performance.

(Reference item b, above.)

O d. Valves 8823 are listed as receiving a close signal. However the P&lD, Figure 6.3-1, shows these valves as closing on a "T" signal instead of an "S" signal. Please explain.

RESPONSE

Note the charging system is not part of the ECCS, and no credit is taken for it in design basis events.

O WAPWR-PSSS 440-51 AMENDMENT 1 2855e:ld MAY, 1986

, - - - - - - - - - m- - -, , - - , , . - , , . - -

a. Module 13 will cover the CVCS. The charging pumps are not loaded onto a safeguards bus upon receipt of an "S" signal, but they are so loaded in the event of a loss of off site power. 'They are not tripped upon an "S" signal, but a "T" signal will cause the isolation of many lines, including the charging lines. A condition of the RCS returning to a pressure above the shutof f (a,c) head of the SI pumps [ [ is not considered to be a problem with the SP/90 because, in part, of the multitude of systems available to reduce pressure, should such an action be needed.

b. The monitor lights will not be used in SP/90, and this information will be removed f rom the SER. An advanced control room will be included in the design. This will be described in Module 16. With respect to the question regarding the valves that are normally open, it is correct that no signal is sent to the valves to open upon receipt of an "S" signal. Indications on the valve positions will be provided to the operator and further, even if one of the valves should be in the wrong position, there are three other SI subsystems which provides more than sufficient I capacity. The advanced control room will further treat this situation, perhaps through something like provision of an overview status indication which indicates that everything is correctly aligned.

c. With respect to valve 9018, note that low pressurizer level will cause isolation of letdown, which in turn will isolate the referenced valve. See also the response to item b, above.

l

d. A "T" signal causes an "S" signal.

440.88 (6.3-4, Section b) This section shows that valves 8871 and 8964 are O 6.3 sent a confirmatory close signal upon Containment Isolation Phase A Signal ("T"). The staf f has not been able to locate ttiese valves in O

440-52 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2855e:ld

the P&ID information provided to the NRC for review. Please identify V the location of these valves in the P&ID information provided to the staff.

RESPONSE

The valves are located on the accumulator P&lD.

440.89 (6.3-4, Section (c)) Unanticipated interactions between various com-6.3 ponents are an item of concern as precursors to accidents. This O. section indicates the opening of containment spray valves upon Containment Phase B Signal ("P"). If the low head pumps are on and injecting water into the RCS (the staff recognizes this is not a normal alignment), the indicated action appears to cause some water to be diverted to the containment sprays.

a. Is this correct?

b. If correct, what is the water flow rate to the containment and to O the RCS under these conditions?

c. If correct, has the reduced flow to the containment been evaluated with respect to the accidents for which the ECCS has been designed and where is this discussed? If not, please discuss.

RESPONSE

a.& This topic was discussed in part in response to 440.77. Note

b. valves 9009 and 9011 must be closed and power removed in order to align for RHR operation (interlock protection). If the situation were to occur, the pump probably would cavitate. It is considered unlikely that one would get into an alignment where runout would be encountered.

c. Question withdrawn by NRC.

O WAPWR-PSSS 440-53 AMENDMENT 1 2855e:ld MAY,1986

440.90 (6.3-5, second paragraph) " Interlocks are provided to automatically 6.3 open the normally closed RHR heat exchanger / component cooling water isolation valves in the event that actuation signals are generated by both the safeguards protection logic ("S" signal) and the high head pump discharge header temperature protection logic (temperature elements TE-866, .... The high head pump head discharge header temperature protection logic consists of one temperature element,

.... A temperature actuation signal would be generated when a temperature channel bistable receives a temperature signal higher than ... (approximately 195'F)."

a. Since the high head pump takes suction f rom the EWST, and the location of the suction line will not be at the top of the EWST, the temperature of the water flowing into the pump may not be representative of the temperature of water flowing into the EWST from containment, nor will it be representative of water temperature in the upper regions of the EWST. This appears to mean cooling water to the RHR heat exchanger is initiated when the pumped water needs cooling, not necessarily when the EWST water needs cooling. Initiation in this manner does not appear to protect the SI pump f rom receiving an overtemperature fluid due to the location of the sensor, the heat exchanger location, and the possibility of significant thermal stratification in the EWST such that a large inventory of hot water may have accumulated prior to action to initiate cooling. This also suggests the possibility that containment spray would occur with hot water rather than with cooled water. Please comment.

Include in the comment a consideration of potential problems of generating a large inventory of hot water in the containment prior to initiation of cooling, if any.

b. It appears that there is only one temperature sensor. Please comment upon what happens if this fails low, so that the indication is that no CCW is needed to the heat exchanger.

I O

WAPWR-PSSS 440-54 AMENDMENT 1 2855e:1d MAY,1986

p c. Are the temperature sensors widely separated as may be implied from the Module 3 layout sketches? Are any portions of the instrumentation system, including wiring or controls, in an area which could be subjected to an adverse environment or are any portions in the same area of the plant? The nature of the concern is if there is a possibility of tenninating CCW to all of the RHR heat exchangers due to a common cause.

RESPONSE

O a. The SI pump is designed to operate with a satisfactory NPSH at a temperature that corresponds to the saturation temperature associated with the maximum expected containment pressure (as determined by conservative methods). Thus, protection of the pump with respect to temperature is not a concern. The SP/90 design does not depend upon containment spray for cooling, but instead uses four safety related containment coolers for that purpose, although, of course, the containment spray will have a cooling effect. Thus, even if the water in the EWST were hot, it would not be a problem. If it were desired to cool the EWST water, suitable flow paths exist so that this can be accomplished

> via the test lines and with the use of the RHR heat exchangers.

The duplication of systems would permit this mode of operation while also operating containment spray, if desirable. Further note, however, that the intent of the containment spray design is

for short tenn operation to reduce any pressure pulse that results f rom a large break, and only a few minutes of operation is necessary to accomplish this.

W does not believe that stratification can occur in the EWST.

The flow rates are high, and slightly less than half of the initial inventory will be held up on the containment floor prior to spilling into the EWST. At this point. the water will have to i fall about ten feet before reaching the water level in the EWST, and this should be sufficient to assure complete mixing. The holdup time for the amount of water on the containment floor is WAPWR-PSSS 440-55 AMENDMENT 1 2855e:1d MAY , 1986

about 20 minutes of full flow containment spray inventory, and this means that for this time span, the water being pumped will be at the initial EWST temperature. Following that time, with complete mixing, slow transients in temperature could occur.

b. Failure of the temperature sensor in this manner would cause water to be boiled on the secondary side of the heat exchanger, with relief through the relief valve. In this respect, the operation is no different than in the existing plants. Note, however, that in the SP/90 there are four systems, and the loss of one is relatively insignificant.

c. Each temperature sensor is located in the corresponding pump compartment, and they are therefore widely separated. No portions of the instrumentation system, including wiring and controls, are located in areas which could be subjected to an adverse environment, and no portions are concentrated in one area of the plant. Note valve position indicators will be provided for the operator and alarms will be provided in the new control monitor system. Further, manual operation of the valves to treat the situation under consideration here will be an option.

l 440.91 (6.3-5, third paragraph) This discusses the interlocks with respect 6.3 to the RHR (low pressure) pump. Please address the issues identified in 440.90a, b, and c (which were with respect to the high head SI pump) with respect to the RHR pump.

RESPONSE

See 440.90.

440.92 (6.3-6, third paragraph) "It is proposed that the four pumping 6.3 modules be housed in containment pressure pump enclosures (CPPE) in order to encompass all piping and components associated with any post-accident recirculation of highly radioactive fluid within a containment boundary." As the staf f has commented previously, the MAPWR-PSSS 440-56 AMENDMENT 1 2855e:ld MAY, 1986

/~ - term " proposed" is not a definitive statement, and a commitment or a clearly defined alternate situation must be provided before the staff can complete review of this design. The staf f in general encourages the use of this concept based upon the limited information which has been provided since it appears to have important advantages in the s post-accident environment. The concept appears to be a direct response to the 10 CFR 50.34(f)(xxvi) requirement for provision for leakage control and detection in the design of systems outside containment that may contain radioactive materials following an accident, and is directly responsive to the goal identified in TMI Action Item 111.D.1.1 to "... minimize potential exposures to workers and public, and to provide reasonable assurance that excessive leakage will not prevent the use of systems needed in an emergency."

It also potentially is in compliance with the NUREG-1070 statement (p.12) that "The commission fully expects that vendors engaged in

~

designing new standard ... plants will achieve a higher standard of severe accident safety performance than their prior designs." In addition, the CPPE concept could effectively reduce the risk

\ associated with event "V" if the CPPE is designed for conditions which could exist for that accident. Please clarify and comment.

(See also 440.29, 440.31, 440.32, 440.33, 440.57b, and 440.74a.)

RESPONSE

See the W response to NRC Question 440.3 in Module 3, as well as 440.29, 440.31, 440.32, 440.33, 440.57, and 440.74.

440.93 (6.3-6, third paragraph) "This total encapsulation concept for the 6.3 ECCS eliminates the potential for post-accident releases of highly radioactive liquid or gases into the auxiliary building and subse-quently into the environment." The staf f would agree the concept appears to significantly reduce the potential, but not that the potential is eliminated. Please provide the bases for your assertion.

RESPONSE

See the W response to NRC Question 440.3 in Module 3.

O AMENDMENT 1 WAPWR-PSSS 440-57 2855e:ld MAY, 1986

i 440.94 (6.3-7, first paragraph) "The four high head pumps are sized such l 6.3 that one high head pump provides sufficient injection flow to prevent core uncovery for small LOCAs up to at least 6 inch diameter break size." Please provide substantiation. Include the transient behavior of vessel level, RCS inventory, break flow rate, pump injection rate, accumulator flow rate, core reflood tank flow rate, '

pressure, pressurizer level, flow rate and velocity through the steam generator tubes, liquid holdup in the steam generator tubes, liquid level in the downcomer, liquid level in the pump suction legs, bypass flow rate between the upper downtomer and the upper vessel, flow rate in the hot legs at the vessel connection, flow rate in the cold legs at the vessel connection, and flow rate in the instrumentation lines used for hot and cold leg temperature measurement (if any). Can H provide best estimate type results as well as licensing type calculations? Please identify which analysis model was used to obtain the ;ults. If the analysis was done with WFLASH, discuss the validity of your conclusion in light of the known nonconservative accumulator injection model.

RESPONSE

See the plots on the following pages. A portion of this discussion, with agreement of NRC, has been postponed until discussion of Chapter 15.

O O

O WAPWR-PSSS 440-58 AMENDMENT 1 2855e:ld MAY, 1986 i

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WAPWR-PSSS 440-66 AMENDMENT 1 2855e:ld MAY, 1986

v 440.95 (6.3-7, third paragraph) "The high head pumps would continue inject-6.3 ing borated water following a steam break until the initial RCS volume had been reestablished with borated water to prevent the possibility of a return to criticality." Please provide the calculated results to substantiate this statement.

RESPONSE

See Module 6/8.

440.96 (6.3-7, last paragraph) "Since the EWST is located inside the con-6.3 tainment, the initial EWST water temperature is approximately 100"F

" Please substantiate this figure for all conditions under which the plant may operate.

RESPONSE

The value is based upon estimates f rom experience gained in other plants. The important point is not '.h e absolute value of the temperature, but the fact that it will not be cold, like of the order

\ of 32"F, and therefore the thermal shock associated with that cold water will not be present.

440.97 (6.3-8, third paragraph) "The four core reflood tanks and the four 6.3 high head pumps provide eight (8) separate means for injecting coolant directly into the reactor vessel. Any combination of five of l

these eight entities are sufficient to meet the large break LOCA functional requirements." Please provide substantiation for this conclusion.

RESPONSE

This conclusion is based on a flow rate which provides one inch per second reflood rate. The design is such that a pump and a CRT deliver at the same rate at the pressure assumed for the analysis.

Note the flow rate needed to keep the core covered for a small break, with the assumption of no accumulator contribution, is roughly half of what would be provided by one CRT at these relatively low pressure conditions.

1 440-67 AMENDMENT 1 WAPWR-PSSS l 2855e:ld MAY, 1986

440.98 (6.3-9, first paragraph) "The component design and operating condi-6.3 tions are specified as the most severe conditions to which each respective component is exposed, during either nonnal plant operation or operation of the ECCS following any postulated event." Please expand upon this and provide a listing of the most severe conditions to which each respective component is considered exposed.

RESPONSE

This material is provided in Subsection 3.9.1 of Module 7,

" Structural / Equipment Design."

440.99 (6.3-9, first paragraph) "By designing the components in accordance 6.3 with applicable codes, and with due consideration for design and operating conditions, the fundamental assurance of structural integrity and operability of the components is maintained." How is this consistent with the staff concern with event "V", which could have essentially been eliminated if the suction portion of the RHR system were designed to the RCS operating pressure? (See 440.31c.)

RESPONSE

As previously discussed, a complete interlock philosophy and removal of power to certain valves is used to prevent the opening of valves to expose the low pressure RHR lines to full RCS system pressure when the plant is at operational conditions. Further, since two valves are provided in series, E considers is very improbable that both could be opened at the same time due to some unforseen situation.

Once the safety considerations have been satisfied, cost becomes a major consideration, and this is what led to the decision to provide a low pressure RHR system.

440.100 (6.3-9 first paragraph) "All discharge piping is water solid during g 6.3 plant operation. Thus water hammers in the injection lines are W precluded." How is this assured? Please include consideration of test operations and elevation effects (such as a head of water in a vertical line with a valve or other component near the top which may g leak air into the piping) in the reply. (See also 440.70) W WAPWR-PSSS 440-68 AMENDMENT 1 2855e:ld HAY, 1986

RESPONSE

Test operations should identify any system leaks so that they can be repaired. The same operations should flush any voids from the system. This should preclude the presence of voids.

O 440.101 (6.3-9. Section 6.3.2.2.1 High Head Pumps) "Each high head pump 6.3 could be either a multistage, vertical or horizontal centrifugal pump driven by an air or water cooled induction motor." Clearly, the pump had not been chosen at the time of preparation of this report. With regard to this pump:

a. What is the present status of pump selection?

b. What is the source of pump characteristic infonnation for the various calculations to support plant behavior predictions?

c. What is the source of the NPSH information previously provided or referenced for the high head pump?

O

d. Please contrast the NPSH requirements for the high head pumps with what is provided in the SP/90 design and discuss the safety margin which exists as a function of plant conditions. The discussion should address the uncertainty associated with the calculations and the pump NPSH requirements as well as fully identifying the assumptions used in the NPSH calculations.

e. If pump selection has not been accomplished, what is the schedule for selection?

RESPONSE

a. The horizontal pump has been selected.

b. Pump characteristics are consistent with the expected performance of domestic pumps, such as furnished'by Bingham or Byron .lackson.

O WAPWR-PSSS 440-69 AMEN 0 MENT 1 2855e:ld MAY, 1986

c. See item b, above.

d. NPSH information provided by pump manufacturers is relatively conservative since this represents one of the specifications which the pump must meet in order to be acceptable. Hence, if the minimum NPSH is met, there is little concern that the pump will operate. In addition, it is felt that minor cavitation will have little effect on pump operation. With this starting point, the excess NPSH provided in the SP/90 configuration is believed to be sufficient.

e. Question withdrawn by NRC.

I 440.102 (6.3-9, Section 6.3.2.2.1 High Head Pumps) "In conjunction with an 6.3 emergency RCS depressurization capability, they provide a bleed and l feed means for emergency core cooling in the event of a loss of secondary side heat removal capability." Please provide the characteristics of operation in this mode, including pressure, heat l

removal rate, heat generation rate in the core, mass rate of addition of fluid to the RCS, mass rate of loss of fluid from the RCS via the valves (PORVs?), and a general description (or reference) of the system and its operation, including any critical timing aspects of placing the system in operation. If the emergency depressurization capability is not capable of a rapid depressurization to below the RHR cut-in pressure, please also address what consideration was given to provision of such a system and why it was rejected. If it is l considered to be capable of a rapid depressurization, provide the timing and other significant conditions necessary to achieve this behavior.

RESPONSE

This was previously discussed in the pre-tendering information on i page 3-77. It will be discussed later in the present effort. Note that information pertinent to the hot leg LOCA will apply to the RCS l depressurization connections to the hot leg. There are no critical l

HAPWR-PSSS 440-70 AMENDMENT 1 2855e:ld MAY, 1986

timing aspects to operation of the depressurization capability.

O Present thinking is that one would not need to use the RHR under conditions suitable for emergency depressurization because it is not needed. Once one has opened up the system to accomplish the depressurization, the high pressure SI pumps, CRTs and accumulators O are more than sufficient to cool the plant,  ;

440.103 One of the potential problems in design of hydraulic systems is pres-6.3 sure relief from pipes which can be isolated, and in which pressure O can subsequently develop due to thermal expansion of the trapped Please describe the W approach and decision process for liquid.

deciding which lines to protect and which do not need protection in the SP/90 design.

RESPONSE

The SP/90 design (ISS) incorporates pressure relief devices to prevent overpressurization of any piping which would be isolated during any normal or expected operations. This includes portions of O piping between closed valves that could be heated as well as portions of the system which are/can be connected to higher pressure systems (e.g. RHR pump suction lines, piping connecting to the high pressure cold / hot leg injection lines to the RCS and the CVCS system interconnections). These relief devices are designed and placed in accordance with ASME Section III requirements.

440.104 (Section 6.3.2.2.1) "These (high head) pumps are protected against 6.3 extremely low flow or no flow operations by a normally open miniflow d path, downstream of each pump, which permits the pump to recirculate a minimum flow back to the EWST, located inside the primary containment building. A containment isolation valve is located in each miniflow path, with power normally removed from the valve to prevent inadvertent closure." Does this mean that containment isolation signals normally have no effect on this valve?

O AMENDMENT 1 WAPWR-PSSS 440-71 MAY, 1986 2855e:1d

. - - - . - ----,-.--,----m,a. ~ _ . - . - - - - - . , , - - - - - - - - , - - - - + - - , , - - -

RESPONSE

That is correct. Containment isolation signals have no effect upon these valves, and they are not intended to be containment isolation valves.

440.105 (6.3-10, Section 6.3.2.2.1) " Analyses are performed to determine the 6.3 net positive suction head (NPSH) available from the EWST. These analyses consider elevation head and piping losses, with no credit taken for containment overpressure. Consequently, this approach meets the regulatory position stated in Regulatory Guide 1.1." The Reg. Guide 1.1 statement is: " Emergency core cooling and containment heat removal systems should be designed so that adequate net positive suction head (NPSH) is provided to system pumps assuming maximum expected temperatures of pumped fluids and no increase in containment pressure from that present prior to postulated loss of coolant accidents." The W statement does not mention temperature, but RG 1.1 specifically identifies temperature. The W statement indicates that analyses are conducted, but there is no statement regarding the results of the analyses and whether these are applied to obtain a design which satisfies the RG 1.1 criteria. There is also no reference to the analyses nor is there a description of their accuracy. Since the W statement is that RG 1.1 is satisfied, please provide a comparison of the RG 1.1 criteria and the specific information showing compliance with the criteria. Include (if not previously provided) the actual elevation of the pump suction as compared to the actual elevation of the bottom of the EWST, a description of the vertical elevations in the suction pipe from the point of connection to the EWST to the pump suction, and the flow rate used for the calculations, in addition to any other information necessary to establish compliance with RG 1.1.

RESPONSE

The W position is that nothing is dm.e here that is any different f rom the approach the vendors use for other plants, and therefore this is a generic item. There is also significantly more NPSH margin O

WAPWR-PSSS 440-72 AMENDMENT 1 2855e:ld MAY, 1986

than shown in the Module 1 submittal. Further, RG 1.1 effectively states that one should not try and force the containment pressure to obtain a needed NPSH.

440.106 (6.3-11, 6.3.2.2.2, Low Head Pumps) Please provide the NPSH informa-O 6.3 tion that establishes compliance with Reg. Guide 1.1. What is the source of the pump characteristic curves when the pump has not been selected? What is the elevation of' the pump suction connection relative to the floor of the room in which the pump is located?

O

RESPONSE

Postponed by NRC. (See response to 440.105).

440.107 (6.3-12, third paragraph) " Water can be removed from the accumulators 6.3 by opening the appropriate valves in the test line and permitting flow to return to the EWST." Please identify the line and valves on the P&ID.

RESPONSE

These are small lines which return flow to the EWST and enter the tank from above. They are not shown on the P&ID. Penetration distance into the tank has not been determined, but there is probably no reason for most lines to extend into the tank for any significant distance. The only lines that connect below the fluid level are the pump suction lines. Purification lines probably will connect to the pump suction lines outside of containment.

440.108 (6.3-12, fourth paragraph) "The margin between the minimum operating 6.3 pressure or level and the maximum operating pressure or level provides a range of acceptable operating conditions. The band widtt is sufficient to minimize the frequency of adjustments in the accumulator pressure or level required to compensate for leakage."

l What are the referenced values? What is the expected leakage rate and path?

lO WAPWR-PSSS 440-73 AMENDMENT 1 2855e:1d MAY,1986

r

RESPONSE

The referenced values are to be generated later in the SP/90 design prog ram. Leakage rates are expected to be small enough to be no problem.

440.109 Page 6.3-12 contains the statement: "A line is provided from the dis-6.3 charge of two high head pumps to the comon accumulator fill line for the purpose of adding water to the accumulators from the EWST while they are at their normal pressure." Page 6.3-13 contains: "A line is provided f rom the discharge of two high head pumps to the comon accumulator fill line for the purpose of adding water to the core reflood tanks from the EWST while they are at their nonnal operating pressure." Page 6.3-7 contains the statement: "There are no piping interconnections between these four separate high head subsystems."

The last statement appears to contradict the former statements.

Please coment. The concern is that interconnections of this type can introduce unexpected systems interactions and statements should not be made that such interconnections do not exist when such is not the situation. In light of the inconsistency, please review the SP/90 design with respect to other similar interconnections, if any, and provide the staff with the results of the review.

RESPONSE

, The line in question is isolated by two normally closed valves and a check valve. One of the valves is also signaled to close upon receipt of a "T" signal. Note that not all pumps have the connection. It is provided to two of them. The statement "There are no piping interconnections between these four separate high head subsystems." is meant with respect to SI usage. It is not considered by W to be an interconnecting line. There are no other similar interconnections. There are small diameter lines which connect to spaces between check valves for test purposes, and these are connected to comon manifolds.

I l

! WAPWR-PSSS 440-74 AMENDMENT 1 2855e:ld MAY, 1986

440.110 ( 6. 3-13) Section 6.3.2.2.3 provides a brief discussion of assumptions v 6.3 used in taking credit for accumulator water during a large or inter-mediate break. There is no corresponding discussion for the core reflood tanks. Please provide the assumptions used for the App. K calculations.

O

RESPONSE

Modeling and assumptions for the CRT's are identical to those used for the accumulators. Note that all four of the CRT's deliver during O. a large break LOCA due to the vessel injection connection which prevents direct spillage out of the break. Water at the end of blowdown is treated in the same manner as for the accumulators; it is thrown away.

440.111 (6.3-14, second paragraph) "The band width is sufficient to minimize the frequency of adjustments in the accumulator pressure or level required to compensate for leakage."

O- a. This is in the section describing the core reflood tanks. Please correct the wording to be applicable to this section.

b. What are the referenced values (levels) and what are the expected sources of leakage and their expected rates?

RESPONSE

a. The sentence should read: "The band width is sufficient to minimize the f requency of adjustments in the core reflood tank (CRT) pressure, or level, required to compensate for leakage."

An amended pg. 6.3-14 is provided in Attachment 440.111.

b. See 440.108.

. O WAPWR-PSSS 440-75 AMENDMENT 1 2855e:1d MAY,1986

440.112 (6.3-14, Section 6.3.2.2.5, Emergency Water Storage Tank (EWST)) '

6.3

a. "Any discharge from the pressurizer relief tank should also be routed to this tank." Has the design determination been made as to whether or not this will be done? Note the F&ID in Section 6.3.

b. "The required water volume depends on the Refueling Canal Volume which is not expected to exceed ... gallons." Is the referenced volume the volume of the EWST or the volume of the refueling canal? Has a final determination been made of the EWST volume and if so, what is it?

c. " Analyses are performed to determine the minimum water level in the EWST during recirculation. These analyses consider the amount of water trapped in lower containment compartments and the delay time for water to return to the EWST." Please provide the results of the analyses. (In particular, include the water holdup volumes and water levels in the various regions of containment).

d. Table 6.3-2 shows a boron concentration in the EWST which does i

not agree with the concentration discussed in respect to other 1

components, such as the accumulators and core reflood tanks.

What is the reason for the difference?

l l

l e. Table 6.3-2 indicates the physical relationship of the EWST to containment. Please describe the methods and equipment used to control contamination of the EWST water due to anything entering l it f rom containment. Also discuss this topic from the viewpoint of holdup of return water flow f rom the containment to the EWST.

f. Please describe all of the lines leading from the EWST that could be used to remove water or could be the cause of water removal.

Note the lines that are identified in the P&lD.

O WAPWR-PSSS 440-76 AMENDMENT 1 2855e:ld MAY, 1986 l

l

'T g. What provisions have been made for cleaning the EWST water during routine plant operations? For sampling? For assuring a homogeneous boron distribution?

h. What consideration has been given to a leak or break in the O EWST7 (Both with respect to the loss of water from the EWST and to the consequences of flooding other portions of the plant.)

1. Please contrast screen total cross sectional area and mesh size O to that typically provided in existing W plants of comparable power level.

j. Please contrast vortexing and air entrainment as potential problems by comparing the SP/90 EWST configuration and the configuration commonly used in existing W plants. Include a discussion of the consideration given to vortexing and air entrainment in the EWST design process. Also address the creation of debris and the approach used to minimize the O potential problem. Further, discuss the plans for procedures to prevent problems and to deal with them if they should occur.

I k. Please contrast the precautions to be taken within containment in the SP/90 to avoid a problem with clogging the EWST inlet with l what has been done in typical existing W plants for containment sumps.

1. Please address material that may cause erosion of the SI and RHR O pumps and valves in the same manner as in item k, above. In addition, please address the concern that has been expressed in dealing with severe accidents which involves debris that may be circulating in the RCS, and hence could enter the EWST, and which O can both erode valve seats and keep valves from operating properly, such as not allowing full closure.

m. Are there any pipe runs in the lines f rom the EWST to the SI and O RHR pumps where water must flow in an upward direction? If so, WAPWR-PSSS 440-77 AMENDMENT 1 MAY, 1986 2855e:ld

'_m

please describe them and address any potential ef fect upon pump performance. The concern is whether there is anything in the pipe configuration where flashing may occur, with an impact upon flow to the pump suction.

n. Are there any connections between the steam generators or steam lines and the EWST? If so, what provisions are made to control potential boron dilution.

o. Please address the influence of channeling due to impingement on the water surface or due to ejection flow from pipes and drains located within the EWST. Include the influence of a high flow l rate f rom the pressurizer relief tank and any steam generator lines, if applicable.

RESPONSE

a. The design determination has been made and the discharge will be routed to the tank.

b. The EWST is larger than the refueling canal. Design iterations l

are still taking place as the design evolves and the plant

! equipment is optimized.

c. Several volumes were provided to the staff. See response to Question 440.236 for a discussion of containment holdup volumes and water level, and EWST level.

d. The difference is to allow for an increase of concentration in l

the accumulators and CRT's during filling.

l l e. Processing capability for cleanup will be provided, probably in Module 13. Several screens will be provided. In general, the j EWST will not be open to containment over its entire surf ace.

I Flow from containment into the EWST will be by way of weirs, with downcomers which penetrate slightly below the normal water level.

WAPWR-PSSS 440-78 AMENDMENT 1 2855e:ld MAY, 1986

l Discussed previously.

s f.

s

g. These have not really been addressed yet. Note that running the test lines should provide mixing. This will be covered in Module 13.

h. The EWST is not considered to break,

i. Total cross sectional area has not yet been determined. The

' selection will comply with the applicable regulations. Mesh size will be comparable to that in existing plants. Trash racks and coarse mesh screens will be provided above the EWST. Fine screens will be provided within the EWST.

j. W will comply with the applicable Regulatory Guides. This is covered in Module 2.

k. There are eight large diameter pipes provided in containment (q./ which lead water into the EWST. Each of the high head SI pumps and each of the RHR pumps has a separate pipe connection to the EWST to supply water for the safeguards operation.

1. Some of this material is discussed in Module 4. The ISSS design l

is such that, in the long term, only a small portion of the l

equipment is needed for cooling. The remainder can be held in reserve and applied if what is being used is worn out due to erosion f rom debris. When the SI pumps are "used up" one can apply each of the RHR pumps in turn, which can also be configured for SI duty. In addition, the separation of components allows one to work on portions of the system while other portions of the system are in use to cool the RCS.

m. There are no such pipe runs.

n. Connections are provided between the steam generators and the EWST. See Module 6/8.

l WAPWR-PSSS 440-79 AMENDMENT 1 2855e:1d MAY,1986

o. Spargers will be used for pipes which connect equipment to the EWST. Flow from containment will be relatively low velocity flow and the impingement on the surface h not expected to cause any difficulty.

440.113 (6.3-15, Section 6.3.2.2.8, Valves) The June 1983 version of this 6.3 document contained the following: " Inadvertent mispositioning of a motor operated valve due to a malfunction in the control circuitry in conjunction with an accident has been analyzed and found not to be credible for consideration in design." This statement is not contained in the more recent version. Please explain.

RESPONSE

The SP/90 PSSS is designed to be unaffected by any single failure, including inadvertent mispositioning of valves.

440.114 Insofar as not provided in the response to 440.85, please provide a 6.3 list of all valves associated with the systems discussed in Section 6.3 which might have their motors (drivers), electrical connections, or controls flooded following an accident. If any are flooded, please provide an evaluation of the consequences for both short and long term ECCS functions. Also list all control room instrumentation l loss following accidents which result in flooding and evaluate the 1

consequences of failures and malf unctions. Further provide similar information, as applicable, with respect to valves for which manual operation would normally be considered as a backup mode of operation, i but which may not be accessible due to flooding.

RESPONSE

i As previously discussed (440.85), this is not considered to be a problem. With respect to the control room instrumentation and manual operation, the intent of the design process is to prevent l dif ficulties of this type. Curbs, flow paths, and elevations, for example, are used to control flooding.

O WAPWR-PSSS 440-80 AMENDMENT 1 2855e:ld MAY, 1986

440.115 (6.3-16, Motor-Operated Gate Valves) Previous discussions in SARs 6.3 submitted to the staf f include " Valves which must function agaiast system pressure are designed such that they function with a pressure drop equal to full system pressure across the valve disc." The wording in the SP/90 submittal under this heading is identical to the O prior submittals, with the exception that the quoted statement is omitted. What is the reason for this change?

RESPONSE

The wording will be restored. An amended pg. 6.3-16 is provided in Attachment 440.115.

440.116 (6.3-17, Manual Globe, Gate and Check Valves) "The stem packing and 6.3 gasket of the stainless steel nanual globe and gate valves larger than 2 inches are similar to those described above for motor-operated I valves." Previous submittals did not contain the qualification with respect to the two inch size. Why?

RESPONSE

There is nothing dif ferent in the SP/90 plant as referenced to other i

W provided plants with respect to this item.

440.117 (6.3-17, Accumulator /CRT Check Valves (Swing-disc))

6.3

a. The CRT valves have to operate with a larger pressure differential as contrasted with the pressure differential for the accumulator valves. Are there any differences in valve design due to this increased operational pressure requirement in comparing the CRT to the accumulator valves?

b. In the discussion of back leakage, the statement is made that "This back-leakage can be checked via the test connection as described in Subsection 6.3.4." This sentence is identical to previous submittals of SERs. However, upon checking in the submitted Section 6.3.4, the referenced material is omitted.

Please explain.

WAPWR-PSSS 440-81 AMENDMENT 1 2855e:ld MAY,1986

c. Prior submittals contain "When the RCS is being pressurized during the normal plant heatup operation, the (accumulator) check valves are tested for leakage as soon as there is a stable l

l differential pressure of about 100 psi or more across the valve.

This test confirms the seating of the disc and whether or not

! there has been an increase in the leakage since the last test."

Does the omission of this information f rom the SP/90 document mean the test is not planned?

RESPONSE

a. The valves are the same as used in other W provided plants, and this is considered to be a generic type question. No differences are provided between the CRT valves and the accumulator valves because all are designed to withstand full RCS system pressure.

b. Several pages of material pertinent to testing were provided in the pre-tendering discussion. See, for example, page 3-80 of the version of Module 1 written in November,1982. This material is applicable to the SP/90.

c. See above.

440.118 A severe ATWS can cause a significant transient pressure rise in the 6.3 RCS. This, in turn, can cause a pressure differential across all valves exposed to the RCS pressure that may be significantly above the design pressure. The major concern is that one or more valves may deform, and then fail to open, thus disabling portions of the SI system such as the accumulators, core reflood tanks, SI injection system, and RHR system. Has this been addressed in the SP/90 design? If so, please describe the consideration given to the potential problem.

RESPONSE

There is no dif ference between the approach used for the SP/90 and other W provided plants with respect to this item. The valves are designated as Class 1 components.

WAPWR-PSSS 440-82 AMENDMENT 1 2855e:1d MAY, 1986

440.119 There are a number of valves associated with the primary side safe-

. (%

6.3 guards systems that are normally in a ready position (either open or closed) for which no movement is necessary if the primary side safeguards equipment is needed. Will the SP/90 have a monitoring system which checks the position of all of the valves and indicates an "0K" status if all are in the proper position, with a corresponding indication and perhaps alarm if any valve is mispositioned? Section 6.3.5.2.6, which provides a discussion of this type of monitor, appears to be specific to " ... valves that are s required to function...." Please discuss.

RESPONSE

Monitoring capability in the advanced control room will continuously evaluate all critical M0V's, whether they are required to move or not.

440.120 Closely related to the above question is the overall question of auto-6.3 matic monitoring of components for which unique lineups are necessary which depend upon the plant operational status, including shut down

( operations, start up operations, and cold shutdown (with such activities as refueling). Is there a plan for coverage of these situations with a single "0K" indication if everything is aligned l properly, and with a suitable indication if a component is not l

l aligned properly or otherwise unavailable? Is there a plan for automatic following of plant response under these offpower conditions and with more specific indication of plant condition tuned to the different needs associated with the different conditions? Again, Section 6.3.5.2.6 touches on this topic, but appears oriented toward power operation or specific phases of the ECCS emergency response as opposed to different normal operational situations of the plant other than power operation. Please discuss.

RESPONSE

With respect to the first half of the item, there is a plan for automatic monitoring with a single "0K" indication. With respect to O

WAPWR-PSSS 440-83 AMENDMENT 1 2855e:ld MAY, 1986

the second half of the item pertaining to the different needs associated with the different conditions, this will be covered in Module 16.

440.121 (6.3-19 Section 6.3.2.5, System Reliability) "The system has been 6.3 designed and proven by analysis to withstand any single credible active failure during injection or any single active or passive failure during recirculation or operator error and maintain the performance objectives. . . ."

a. Has the analysis referred to above been completed?

b. In general, what is the W position with respect to active vs.

passive failures and when is one to be considered vs. allowing for both? What types of passive failure are considered and what types are not?

c. What is the definition of " operator error" as used in the above?

Will this definition apply to all of the SP/90 documentation?

RESPONSE

a. Yes. It is presented in Chapter 15.

b. This question is considered to be of a generic type. In general, the classic definitions and practice of past W designed plants l

are applicable. With respect to the PSSS and the SP/90 plant, a failure can be tolerated regardless of timing.

Operator error means an operator can make a single incorrect or O

c.

omitted action. This definition will apply to all aspects of the SP/90 program.

O 440.122 (6.3-20, second paragraph) "The preoperational testing program 6.3 ensures that the systems as designed and constructed will meet the functional requirements as calculated in design." Is something left out of this sentence and, if so, what?

MAPWR-PSSS 440-84 AMENDMENT 1 2855e:ld MAY , 1986

RESPONSE

O, The sentence should read: "The preoperational testing program

ensures that systems, as designed and constructed, will meet the functional requirements." An amended pg. 6.3-20 is provided as p Attachment 440.122.

V 440.123 (6.3-20, third paragraph) "

... only designs which have been proven by l 6.3 past use in similar applications are acceptable for use...." as ECCS t

components. Does this preclude the improvement of recognized faults l

in components? Does the EWST fit within the limits imposed by this statement? What is the past use with respect to the SI and RHR pump configurations? If the " pump houses" connect to containment under j some accident conditions, where is the experience applicable to those connections and their operation? The concern is that sufficient flexibility be allowed so that improvements important to the reduction of risk can be fully considered (while simultaneously, the staff recognizes the importance of experience with equipment as a factor in reliability).

V

RESPONSE

The intent of the statement is that known f aulty designs are not

! incorporated into the SP/90, and this is accomplished in general by using components which have been proven in practice. The EWST is considered to be a proven component in that only the shape is new.

Many tanks of this general type have been built in the past. With l

respect to the pump houses and connections to containment, this

! concept is no longer a part of the SP/90 design, as previously discussed, and therefore the connection is not applicable to the discussion.

440.124 (6.3-21, first paragraph) This paragraph, which introduces the fail-

! 6.3 ure mode and effects analysis, does not identify the analysis as specific to LOCA, nor does the referenced Table 6.3-4 contain a clear definition that this is the case. Is the staff correct in assuming O

WAPWR-PSSS 440-85 AMENDMENT 1 2855e:1d MAY , 1986 l

this to be LOCA specific (although perhaps applicable to steam generator tube rupture and secondary side rupture in the short term due to similar demands placed upon the ECCS)?

RESPONSE

e staff is correct.

440.125 (6.3-21, first and second paragraphs) The first paragraph, in des-6.3 cribing the ECCS performance with an active failure, ends with

...and still meet the required level of performance for core cooling." The second paragraph, in discussing the applicability of the failure mode and effects analysis to steam generator tube rupture and steam generator secondary side rupture, ends with ". . .and still meet the level of performance for the addition of shutdown reactivity." Does this mean that the only item being addressed in the second case is for shutdown reactivity and that all of the other aspects of the ECCS for those accidents are not being addressed in this section of the SP/90 documentation?

O

RESPONSE

No. Aspects other than reactivity are considered.

440.126 (6.3-20, 6.3-21) The discussion of active and passive failure cri-6.3 teria does not mention those systems which provide services required for the operation of the ECCS. Please discuss or provide a specific reference to where the information will be found for all of the services required for the operation of the ECCS.

RESPONSE

This material has been (or will be) covered, for the most part, in Module 9. " Instrumentation & Controls and Electric Power," and Module 13, " Auxiliary Systems," which provide discussion on the applicable service systems.

I 1

440-86 AMENDHENT 1 e

i WAPWR-PSSS 2855e:ld MAY,1986 l

l l

440.127 (6.3-22, first paragraph) "Any post accident recirculation leakage 6.3 outside the containment building, would be contained within the individual safeguard component areas."

a. What is the pressure containment capability of the housing which contains each of the four ECCS subsystems with its associated pumps, valves, piping and containment penetrations?

b. What leak rate can be tolerated within a housing and have the subsystem remain in operation?

c. What leak rate can be tolerated within a housing with the subsystem no longer operative?

d. Can a housing be isolated completely so that a leak within a housing will be contained as long as the pressure remains within the pressure as identified in item a, above? At some other pressure?

RESPONSE

As previously discussed, this concept is no longer a part of the SP/90 design. With respect to item a, the pressure capability was to have been the same as containment.

440.128 (6.3-22, third paragraph) "The proposed containment pressure pump 6.3 enclosures (CPPEs) are designed to perform all the functions of the safeguard component area with an additional internal design pressure i O requirement which is consistent with the containment design pressure."

i

a. The staff has previously addressed the wording " proposed" in that i it is not definitive.

I

b. What is the design pressure of the CPPE and what is the design 1 pressure of the containment?

O WAPWR-PSSS 440-87 AMENDMENT 1 2855e:ld MAY, 1986

c. What are the anticipated failure pressures of each of the above (item b)?

RESPONSE

See response to 440.127.

440.129 (6.3-22, fourth paragraph) "In the event of leakage inside the CPPE, 5.3 the sump pumps would automatically start and pump the leakage back into the in-containment EWST. Redundant level, pressure and radiation instrumentation would provide the operator with precise diagnostic information with regard to the quantity of leakage and/or leakage rate."

a. What is the capacity of each of the sump pumps?

t

b. How does one obtain flow rate information f rom the above listed instrumentation?

c. Where are the described connections shown on the P&ID which l illustrate the connections to the EWST?

d. What other means are provided, if any, for the return of fluid f rom the CPPE to the EWST? If any, please describe including l design flow rates and the conditions applicable to the determination of those flow rates, and the means of isolating the CPPEs from the containment. Include actions which occur upon encountering a containment isolation signal. Also include a l

description of the conditions under which the containment may no longer be isolated with respect to the common connection with the l CPPE, if such a connection exists. Please provide the background information applicable to the statement on page 6.3-23: "The CPPEs are designed to totally contain highly improbable nondesign base failures such as a major pipe rupture of any post accident or normal cooldown recirculation piping."

O WAPWR-PSSS 440-88 AMENDMENT 1 2855e:1d MAY, 1986 l

e. Are any means taken for purification of sump water prior to return to the EWST7 If so, please describe, including the capacity of the treatment system and a description of operation if the capacity is exceeded.

O f. How is the atmosphere in the CPPEs controlled? Include consideration of both thermal and radioactive release.

RESPONSE: _ _

a. Each sump pump will have a capacity of ~

GPM. UnderI CI normal conditions, the liquid will be pumhd to normal auxiliary building handling systems. Under accident conditions, the option will exist to return the flow to the EKST.

b. Question withdrawn by NRC.

c. As previously discussed, only the major lines connecting between the PSSS and the EWST are shown on the P&ID.

d. As previously discussed, the CPPE concept is not part of the SP/90 design,

e. No purification of water is performed prior to return to the EWST.

l f. The original thinking was that each of the pump compartments would be a sealed room with a room cooler for temperature control and a purge capability for atmospheric processing.

440.130 (6.3-23,second paragraph) "The provisions for leakage detection 6.3 within the ECCS meet the recommendations of NUREG-0737, Item 111.0.1.1." Please list each of the recommendations of NUREG-0737, O item 111.D.1.1, and show how each is met by the SP/90 design.

RESPONSE

This is provided in Section 5.2.5 of Module 4.

440-89 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2855e:ld

9 440.131 (6.3-25, first paragraph) "This signal would automatically realign 6.3 the suction of the charging pumps from the volume control tank to the spent fuel pit.

a. What is the boron concentration of the water in the spent fuel pit and how is this concentration assured?

b. What are the elevations of the connection to the spent fuel pit and of the pump suction connection?

c. What is the NPSH available as contrasted to that required under these conditions? Please also address the water level in the spent fuel pit and the control of that water level to assure sufficient inventory for the above described operations with the charging pumps.

RESPONSE

a. The concentration is ppm. Further information will be provided in Module 13.

i b. The connections will be within about a foot of the top.

l

c. The charging pumps are located near the bottom of the layout and I there is no difficulty with NPSH because of the elevation provided.

440.132 (6.3-25, third paragraph) "The operator should therefore use the l 6.3 charging system to its maximum capability, if required, to maintain water in the pres surize r. " Does this include the positive displacement pump and if so, do the statements with regard to the l realignment of the pump suction also apply to the positive displacement pump?

, RESPONSE:

The positive displacement pump is not included with regard to this discussion.

WAPWR-PSSS 440-90 AMEN 0 MENT 1 2855e:ld MAY,1986

'p 440.133 (6.3-25, second paragraph) "Following this type of small RCS leak, 6.3 ....The operator should therefore use the charging system to its maximum capability, if required, to maintain water in the pres suri ze r. "

O

a. What is the anticipated amount of water that would be added to the containment following this procedure for a small break inside containment?

, b. What water level will this cause inside containment?

i c. What are the implications of this level inside containment, if any, and would it be better to use the SI system which would draw water from containment?

RESPONSE

a. On the order of 100,000 gallons. ,

b. Question withdrawn by NRC.

c. Question withdrawn by NRC. ,

440.134 (6.3-26, top of page) This switchover procedure from hot leg to cold 6.3 leg injection appears to result in the complete removal of injection to the RCS prior to initiation of hot leg injection. If this is correct, why was this route selected as contrasted to one in which RCS injection would be maintained?

RESPONSE

This is not correct. The switchover would take place one subsystem at a time, and injection therefore would not be terminated.

440.135 (6.3-26 third paragraph) "Af ter several hours of hot leg injection; I 6.3 one or more of the high head pumps would be realigned to deliver directly to the reactor vessel injection nozzles, thus establishing a 440-91 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY , 1986

---.-.-n------ -

,-,----____-.--.n--- ,-,,,---,.-,----.-,--m,--,.,-------,- -

en ,, . - , , , - , -n - , , - , , - - - - - , ,

simultaneous flow to the reactor vessel downcomer and the RCS hot legs." This process, as described in the switchover procedure, has the operator first going completely to hot leg injection from cold leg injection, and then switching partially back. Why was this )

selected as contrasted to only switching partially to hot leg injection and leaving a portion of the cold leg injection alone, thereby reducing the number of operator actions as well as reducing the number of operations required of the equipment?

RESPONSE

The selected procedure is felt to be simpler on an overall basis due to the exclusion of branch paths in the procedures that would otherwise be necessary.

440.136 The staff position concerning boron dilution has been previously out-  !

6.3 lined as follows (See, for example, 0 440.33 for Millstone Nuclear Power Station Unit No. 3):

a. The boron dilution function shall not be vulnerable to a single active or limited passive failure (i.e., leakages of seals).

Specifically, the limiting single active failure should be considered during the short-term period of cooling. During the long-term period of cooling, the limiting single active f ailure should be considered and so should a limited passive failure be considered, but not necessarily in conjunction with each other.

i

b. The inadvertent operation of any motor operated valve (open or closed) shall not compromise the boron dilution function, nor shall it jeopardize the ability to remove decay heat f rom the l primary system.

l

c. All components of the system which are within containment shall ,

be designed to Seismic Category I requirements and classified Quality Group B.

O WAPWR-PSSS 440-92 AMENDMENT 1 2854e:ld MAY, 1986 l

h i

d. The primary mode for maintaining acceptable levels of boron in the vessel should be established. Should a single failure disable the primary mode, certain manual actions outside the control room may be allowed, depending on the nature of the action and the time available to establish the backup mode.

e. The average boric acid concentration in any region of the reactor vessel should not exceed a level of four weight percent below the solubility limits at the temperature of the solution.

O the ECCS normally l

f. During the post-LOCA long-term cooling, operates in two modes: the initial cold leg injection mode, followed by the dilution mode. The actual operating time in the cold leg injection mode will depend on plant design and steam binding considerations, but in general, the switchover to the dilution mode should be made between 12 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after LOCA. ,

t fp g. The minimum ECCS flow rate delivered to the vessel during the dilution mode shall be sufficient to accommodate the boil-off due to fission product decay heat and possible liquid entrainment in the steam discharged to the containment and still provide sufficient liquid flow through the core to prevent further increases in boric acid concentration.

h. All dilution modes shall maintain testability comparable to other ECCS modes of operation (HPI-short term, LPI-short term, etc.).

The current criteria for levels of ECCS testability shall be used as guidelines (i.e., RG 1.68, 1.79, GDC 37).

Please describe the ability of the SP/90 design to comply with this position.

RESPONSE

The SP/90 complies with the above.

. O WAPWR-PSSS 440-93 AMENDMENT 1 2854e:1d MAY,1986

440.137 (6.3-27, last paragraph) Please contrast the 2500 ppm boron in this 6.3 statement to the concentrations given in Table 6.3-2 and explain the differences.

RESPONSE

The difference is provided so that adjustments of the boron concentration can be made.

440.138 (6.3-27, next to last paragraph) " Safety injection signal actuation 6.3 will rapidly close all feedwater control valves and feedwater isolation valves, trip the main feedwater pumps, and close the feedwater pump discharge valves." What actions occur with respect to the emergency feedwater?

RESPONSE

A start signal is sent with respect to the emergency feedwater. This was discussed on page 3-70 of the pre-tendering submittal, and is covered in Module 6/8. See also the response to 440.142, 440.139 (6.3-27, item 3) The list of Chapter 15 accidents that result in or 6.3 are affected by ECCS operation include "3. Spectrum of rod cluster control assembly (RCCA) ejection accidents". All of the other items i in the list are discussed in the paragraphs which follow except the quoted one. Where is that discussion located?

RESPONSE

The spectrum of rod ejection accidents are analyzed, from a radiological standpoint, in Chapter 15. The consequences of a small break LOCA in the vessel upper head, due to a rod ejection, are bounded by the small breaks presented in the FSAR.

440.140 Section 6.3.3.1 addresses failure of a single steam dump, relief, or 6.3 safety valve. Is there a common cause failure mode which could cause more than one valve to fail open at the same time?

O WAPWR-PSSS 440-94 AMENDMENT 1 2854e:ld MAY , 1986

l I '

RESPONSE

The question is applicable to Module 6/8.

440.141 (6.3-28, third paragraph) "A makeup flow rate from one nonnal CVCS 6.3 charging pump is more than adec,uate to sustain pressurizer pressure O at 2,250 psia for a break of a ... inch diameter. This size break corresponds to a loss of RCS inventory of approximately ...." Please compare the loss rate to the makeup rate from one normal CVCS charging pump. The loss rate calculation should be performed for the O expected conditior.s at the break and the charging pump rate that should be used is the condition cxisting at the pump.

RESPONSE

The request is applicable to Module 13.

l 440.142 (6.3-28, last paragraph) "The "S" signal stops normal feedwater flow 6.3 by closing the main feedwater line isolation valves and initiates

, p emergeni.y feedwater flow." Has any consideration been given to also l initiating startup feedwater flow, or perhaps to initiating startup feedwater flow followed by emergency feedwater flow if the startup feedwater flow is inadequate?

i

RESPONSE

The startup feedwater flow pump is not automatically connected to the emergency diesel. See Module 6/8.

440.143 (6.3-29, first paragraph) " ...one high head pump provides sufficient O 6.3 injection flow to prevent core uncovery for small LOCAs up to a 6.0 inch break size." Please provide the results of the calculation as requested in 440.94.

RESPONSE

See the response to 440.94.

1 O

WAPWR-PSSS 440-95 AMENDMENT 1 2854e:ld .

MAY,1986

440.144 (6.3-30, first paragraph) In discussing the break behavior modeling 6.3 to meet the requirements of App. K, the following statement is made:

"The conservative assumption is made that the water injected (from the accumulators and the core reflood tanks) bypasses the core and goes out through the break until the expulsion or entrainment l mechanisms responsible for the bypassing are calculated not to be effective." Is the described approach a previously approved method j and is it consistent with the meaning provided in the second paragraph of page 6.3-307 If not, please discuss the expulsion or entrainment mechanisms and the means used to calculate that they are no longer effective. )

i

RESPONSE

The approach is consistent with both the previously approved method and the second paragraph of page 6.3-30.

440.145 (6.3-31, items a - d) These are not precisely the criteria defined in 6.3 10CFR50.46(b). For example, in item c. the statement is "the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling" whereas the comparable statement in 10CFR50.46(b) is " Calculated changes in core geometry shall be such that the core remains amenable to cooling. The W statement in item

d. is "The core temperature is reduced and ....

whereas 10CFR50:46(b) requires "

. . .the calculated core temperature shall be maintained at an acceptab y low value ....

Please explain the differences in wording ars 'jhe implications and provide a precise statement covering compliance with the requirements or deviations that are being requested.

RESPONSE

This is standard boiler plate in H FSAR's. W does not take exception to any Appendix K requirements. Since there seems to be some confusion, the wording will be changed to that contained in Subsection 15.6.4.1 items (d) and (e) of this module. An amended pg.

6.3-31 is provided.

WAPWR-PSSS 440-96 AMENDMENT 1 2854e:ld MAY, 1986

440.146 (6.3-31, second paragraph) In the discussion of plant response to a 6.3 steam line break, the statement is made that "A return to power following a steam pipe rupture is a potential problem; however, analysis demonstrates that the core is ultimately shut down by the injected boric acid." On page 6.3-7, the statement is made O "

... prevent the possibility of a return to criticality." What is the situation with return to criticality following a steam line break?

N RESPONSE:

See Module 6/8.

440.147 (6.3-31, third paragraph) "The water initially within the high head -

6.3 pump piping is assumed to be swept into the RCS (with no credit taken for its boron) befoce the 2500 ppm water f rom the EWST reaches the core." What is the reason for this assumption?

RESPONSE

The only reason is to be conservative.

440.148 (6.3-32, item c.) Please provide the overtemperature information 6.3 which leads to reactor trip that is referenced in this item.

RESPONSE

Item c should read as follows: "c. Continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure or low DNBR reactor trip. The "S" signal automatically terminates normal feedwater supply and initiates emergency feedwater O, addition. Af ter reactor trip, the break flow reaches equilibrium at the point where incoming safety injection flow is balanced by outgoing break flow. The resultant break flow persists from plant trip until operator action is taken to bring the primary system and O the faulted steam generator secondary system pressures into equilibrium." The corrected text is given in Attachemnt 440.148.

O WAPWR-PSSS 440-97 AMENDMENT 1 2854e:ld MAY, 1986

440.149 (6.3-35, first paragraph) In regard to the response of the plant to a 6.3 spurious SI:

a. Please provide a transient RCS pressure curve and a transient pressurizer level curve for a normal reactor trip.

b. What occurs with the charging pumps under a spurious SI condition?

c. What occurs if an automatic reactor trip is not obtained if there ,

is a spurious SI? Include operator response actions.

RESPONSE

a. Reactor trip events are performed as part of system design transients. The severity of the cooldown transient follcwing a reactor trip depends upon the assumptions made with regard to secondary side cooling. Based on the results of the APWR reactor trip transients the RCS measure can--but does not always--fall below the shutoff head of the safety injection pumps. Since the actuation of safety injection following a reactor trip is not a safety issue it is not felt that these plots need to be provided

at this time. However, the writeup should be revised as shown in Attachment 440.150(a).

b. The charging pumps would continue to run following an SI signal if offsite power is available.

l l

c. Several scenarios are possible if a reactor trip is not obtained from a spurious "S" signal.

1. No reactor trip, no EFWS, no feedline isolation. Plant just sits there since shutoff head of SI pump is well below operating pressure.

2. No reactor trip, no EFWS, feedline isolation. Loss of feedwater transient ensues. Results are covered by Chapter 15 loss of feedwater flow.

WAPWR-PSSS 44o_98 AMENDMENT 1 2854e:ld MAY,1986

( 3. No reactor trip. EFWS actuation, feedwater isolation.

Results less severe than loss of feedwater results presented in Chapter 15.

Specific operator actions are not yet available for APWR.

Actions are expected to be similar to those for current plants with low level SI pumps.

440.150 (6.3-35, Section 6.3.4.1 ECCS Performance Tests) 6.3

a. The first paragraph contains: " Preliminary operational testing of the ECCS can be conducted during the hot functional test of the RCS following flushing and hydrostatic testing, with the system cold and the reactor vessel head removed." Is something left out of this statement? It appears to be contradictory.

b. It is not clear from the discussion whether the tests are conducted so that the diesels are loaded from all plant loads or just from the SI. Please clarify.

c. The last paragraph of this section contains "Each accumulator and core reflood tank is filled with water from the and pressurized with nitrogen with the motor-operated valve on the discharge line closed." Please clarify by supplying the missing portion of the sentence.

d. There is no mention of RG 1.79, "Preoperational Testing of ON Emergency Core Cooling Systems for Pressurized Water Reactors,"

in this discussion. Please provide a listing of each of the items specified in RG 1.79 and show the extent to which the SP/90 design is in compliance.

O

e. There are numerous check valves and some isolation valves (such as the accumulator and core reflood tank lines) for which no test O

WAPWR-PSSS 440-99 AMENDMENT 1 l 2854e:ld MAY, 1986 l

1

sequence is mentioned. What is to be accomplished to assure proper operability of these valves (which are not exercised as part of the testing with return flow into the EWST)?

RESPONSE

a. The correct wording is: " Preliminary operational testing of the ECCS can be conducted following flushing and hydrostatic testing, with the system cold and the reactor vessel head removed." An amended pg. 6.3-35 is provided in Attachment 440.150(a).

O

b. This is applicable to Module 14 or to the Technical Specifications.

c. The wording should be "Each accumulator and core reflood tank is filled with water from the EWST and pressurized with nitrogen with the motor-operated valve on the discharge line closed." An amended pg. 6.3-36 is provided in Attachment 440.150( c ) .

d. NRC's position on this Reg. Guide has been that showing the applicable tests is sufficient. This will be covered in an appendix to Module 2. Reference is also made to page 1.8-12.

e. Some of the previous discussion on related topics is applicable l here. The real concerns are failure to reclose or parts breaking off and blocking flow types of incident's. Important to these considerations is the SP/90 configuration. In today's plants, all three of the accumulators have to work with a break of significant size, for example. This is not the situation on the l SP/90 on the basis of best estimate calculations, although this tentative conclusion needs to be verified.

440.151 (6.3-38, top list of signals used to initiate the "S" signal) Item d 6.3 states " Excessive cooldown, produced by low T-cold signals in one loop coincident ....

How many signals are sampled and what is the rational for acceptance with respect to the "S" signal?

WAPWR-PSSS 440-100 AMENDMENT 1 2854e:ld MAY, 1986

i

! RESPONSE:

The text must be revised since the protection logic has been changed. Current logic uses one T signal f rom each loop with 2 ,

cold out of 4 coincidence to initiate safety injection. An amended pg.

6.3-38 is provided in Attachment 440.151.

440.152 (6.3-38) This discussion of containment isolation signals contains no 6.3 mention of radiation level. Was this source of information considered as a portion of the containment isolation process? Please discuss.

RESPONSE

< It was not considered in this module. W does not use radiation level for generation of a containment isolation signal. See SRP 6.2.4 and also II.E.4.2.(7) of the TMI Action Plan.

1

> 440.153 (6.3-39, item a.) What is the physical location of the pressure indi-6.3 cator in the high head pump suction line with respect to the pump inlet and the inlet elevation? Is a signal available that could be ,

' used for readout at any location other than the physical location of  ;

the pressure indicator?

! RESPONSE:

This is a local readout indication only and is located close to the pump suction.

440.154 (6.3-39, item c) Please respond to 440.153 with respect to the RHR O 6.3 pumps.

RESPONSE

l See 440.153.

O WAPWR-PSSS 440-101 AMENDMENT 1 2854e:ld MAY, 1986 i

440.155 (6.3-39, item e) In regard to the check valve test-line header pres-6.3 sure;

a. What is the physical location of this instrument?

b. Can this local reading indicator be used under operational or accident conditions?

RESPONSE

a. This is located outside of containment.

b. No.

440.156 (6.3-40, items f and g) What is the amount of deviation that is con-6.3 sidered acceptable prior to alann on accumulator and CRT pressure (or where will this information be defined)?

RESPONSE

See prior discussion. The set points will be selected so that adequate consideration is given to this point.

440.157 (6.3-40, last item) A low flow rate alarm is indicated on high head 6.3 pump miniflow. What is the flow rate in the miniflow line as a function of RCS pressure and will the indicated alarm be actuated if the RCS pressure is low?

RESPONSE

(NRC reworded this question to "A low flow rate alann is indicated on high head pump miniflow. Will the alarm be actuated if the RCS l

pressure is low?)

Yes. W will reconsider this alarm with respect to low pressure behavior.

O WAPWR-PSSS 440-102 AMENDMENT 1 2854e:ld MAY, 1986

440.158 (6.3-41, item c) In regard to the low head pump flow element and

6.3 transmitter

a. What is the rational that will be applied to determination or low flow criteria and the low flow rate alarm? (What will prevent alarms when pressure is above the shutoff head of the pump? What will assure an alarm if the flow rate is less than required for each of the operational alignments?)

b. What is the rational behind not providing a low flow alarm for the high head pump and providing one here?

RESPONSE

a. The low flow alarm is intended to alert the operator that the flow from an RHR/CS spray pump during the RHR operating mode is less than or has decreased below the expected flowrate. Since the post-accident mode of operation for the RHR/CS pump is containment spray (not core cooling) and since the containment spray flowrate is greater than the flowrate provided for RHR operation, no alarm will be activated.

b. The high head pump is provided with a miniflow rate alarm which eliminates the need for a flow rate alarm.

440.159 (6.3-41, item e) In regard to the connections and flow readout indi-6.3 cations involving the test-line header:

a. The staf f is unable to determine where the flow terminates af ter leaving the instruments in question. Please provide this information or discuss where it can be found in information already submitted.

O

b. What is the design pressure of the line containing this instrumentation?

c. What is the physical location of this instrumentation?

WAPWR-PSSS 440-103 AMENDMENT 1 2854e:ld MAY, 1986

d. Can this local reading be readily used under operational or accident conditions?

e. The staff has been unable to identify the connections between this instrumentation and the CRTs. Please provide this information.

RESPONSE

a. The flow terminates in the EWST.

b. This is a 2501 line which will withstand RCS operational pressure.

c. This is located outside of containment.

d. No.

l

e. This is shown on the P&ID.

440.160 (6.3-41, last paragraph) In the discussion of CCW flow to the RHR l 6.3 heat exchangers, reference is made that flow is initiated only if actuation signals are generated by both the safeguards protection logic and the high head pump discharge temperature protection logic.

l What happens with respect to CCW if the "S" signal is reset?

RESPONSE

i l There is no effect and no change.

440.161 (6.3-42, item b on low head pump discharge header temperature) A tem-6.3 perature element is referenced in the discharge header of each low head pump which also represents the inlet temperature to each RHR heat exchanger. The statement is made that "The automatic opening of the RHR heat exchanger / component cooling water isolation valves would be initiated in the event that actuation signals were generated by the RHR pump discharge temperature protection logic." But the pump inlet temperature is in part controlled by the bypass flow when WAPWR-PSSS 440-104 AMEN 0 MENT 1 2854e:1d MAY, 1986

overall flow through the pump is low. What is the rational behind this statement in regard to pump protection? See also 440.90 and

! 440.162 in regard to this method of control as contrasted to the desired cooling in the containment. The staff concern is that the O inlet temperature to the pump will not respond to the actuation i

signal for some time due to the flow path, the large volume of water in the EWST, and thermal stratification and separation which may occur.  ;

RESPONSE

The RHR/CS pump miniflow heat exchanger assures that sufficient cooling of the pumped fluid is provided when the net pump flow is low. The temperature channel at issue does not provide protection j for the RHR/CS pump. The reference to protection logic refers to the fact that the initiation of CCW flow to the RHR heat exchanger is a safeguards function during the long term recirculation mode (HHSI pump discharge temperature channel TICA-866). The temperature i -

channels (TICA-912, 913, 914, 915) in the RHR/CS pump discharge

! piping to the RHR heat exchanger assure that CCW flow also is automatically initiated during " warmup" of the RHR portion of the ISS prior to RCS RHR cooldown.

The quoted sentence should read "The automatic opening of the RHR heat exchanger / component cooling water isolation valves would be initiated in the event that actuation signals were generated by the RHR pump discharge temperature logic." An amended pg. 6.3-42 is provided in Attachment 440.161.

i 440.162 (6.3-42, item c) The RHR heat exchanger outlet temperature is located 6.3 as identified, downstream of the flow bypass return: This would appear to mean that for conditions of no flow in the line where the ,

sensor is located, such as may occur if flow is being bypassed, that there is no correlation between RHR outlet temperature and the temperature indicated. Further, there may not necessarily be a i

WAPWR-PSSS 440-105 AMENDMENT 1 i 2854e:1d MAY,1986 i

^

correlation between the main control board indication of delta temperature of RHR flow and what is actually occurring with respect to the RHR heat exchanger. Please comment.

RESPONSE

(The staff deleted the sentence pertaining to the correlation between the main control board and what is actually taking place.)

Temperature indication is provided on the CCW system on the miniflow heat exchanger.

440.163 (6.3-42, item c) In regard to the local indication of RHR heat exchanger temperature in the outlet line immediately downstream of the heat exchanger:

a. Is there any signal associated with this sensor that is available outside of containment?

b. The statement is made that: "It provides a means of performance verification and heat balance when used in conjunction with the RHR heat exchanger inlet temperature indication of TE-912 ...."

Although this is correct with the proper valve lineup for operation of the RHR pump, it may not be correct under other conditions. Should a qualifier be added? And why was a similar statement not made with respect to TE-866 which is in the outlet of the high head pump?

RESPONSE

a. No.

b. The application is for preoperational testing and checkout. W opinion is that no qualifier is needed. And with the high head pump, the statement was not provided because the heat exchanger could be verified with the instrumentation provided, as discussed above.

O WAPWR-PSSS 440-106 AMENDMENT 1 2854e:ld MAY, 1986 1

440.164 (6.3-43, first item) "A minimum of four level instruments are recom-6.3 mended f or the EWST. . . ." How many will be provided or when and how will this determination be made? How many out of how many will be used for the indicated high and low level alarms? (See also 440.165 where the referenced discussion indicates the selection appears to have been made.)

RESPONSE

O Four will be provided. With respect to the selection, see 440.165b.

V This is correct as stated there.

(6.3-43, last paragraph) The Low-Low EWST level alarm setpoint is 440.165 6.3 identified as based on a minimum allowable level in the EWST to ensure that adequate NPSH is available for the high head and the low head pump operation.

a. "This alarm would be produced on receipt of the EWST low level actuation signal." is this related to the low level alarm and j

i level? If so, how?

b. "The EWST low level actuation signal would be initiated when two out of four EWST LO-LO level channel bistables received a EWST level signal lower than the predetermined LO-LO level l

setpoint.... each level channel is assigned to a separate vital instrument bus." What is the behavior of these buses with loss of power conditions and how is the two of four logic affected?

c. Is there any relationship between the EWST level indications, levels, and the RG 1.82 specifications? Please discuss.

d. What is the time difference between low-low level alarm and the O- time when adequate NPSH is no longer available? Please include consideration of instrument error under normal and accident conditions in the response.

440-107 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:ld l

RESPONSE

a. They are the same.

b. These are loaded onto the diesels and the behavior is unaffected by loss of power.

c. The SP/90 Will completely meet or exceed the RG 1.82 specifications.

d. This is not applicable to the SP/90 since inadequate NPSH cannot occur.

440.166 Alarms are provided on many of the tanks, such as the accumulators, 6.3 core reflood tanks, and the EWST. The setpoints of these ordinarily will be exceeded during an expected response to some operational and accident conditions. Does this mean that the alarms on the SP/90 will sound for these expected conditions if they occur?

RESPONSE

This should be addressed in future modules.

440.167 (6.3-44, section 6.3.5.2.5, fif th line) "... valves is not fully open 6.3 RCS pressure above the safety injection unblocking..." Is something missing?

l

RESPONSE

This should read "

... valves is not fully open at an RCS pressure above the safety injection unblocking..."

( 440.168 ( 6. 3-45) This page contains a generalized description of a series of 6.3 light boxes so arranged that the contained lights are either all on or all off, depending upon the condition of the plant, for the components represented within the boxes. The idea presented is that any light not in phase with the other lights within a box that represents the operational state of the plant would immediately identify a potential problem.

WAPWR-PSSS 440-108 AMENDMENT 1 2854e:ld MAY, 1986

- ' a. The discussion is generalized and not specific. When will a specific description be available?

b. What consideration was given to alternate schemes based on state of the art human factor engineering? Please discuss.

RESPONSE

This subject will be covered in Module 15, "ACR/ Human Factors," which addresses the Advanced Control Room (ACR).

O 440.169 (6.3-47) This table identifies two valves for EWST makeup from the 6.3 CVCS, two valves for EWST supply to the SFPCS, and two valves for EWST return f rom the SFPCS. Where are these valves located on the P&ID?

RESPONSE

The piping connections to and f rom the EWST which provide for level

! control, boron concentration control, and purification will be shown l as part of the Spent Fuel Pit Cooling and purification system. This information will be included in the Auxiliary Systems Module, No.

13. These piping connections will all include " fail closed" isolation valves. Table 6. 3-1 will be updated as part of the Integrated PDA submittal.

440.170 ( 6.3-49 ) What is the meaning of the notation " Spherical containment 6.3 only" since this is the design that is being presented and to the staff's knowledge there is no other encompassed in the SP/90 documentation that is under review?

l

RESPONSE

The correction is made in Attachment 440.170.

O WAPWR-PSSS 440-109 AMENDMENT 1 2854e:ld MAY, 1986 l

l

440.171 (6.3-50) 6.3

a. What is the water volume in the accumulators and the core reflood tanks that is expected to be discharged during a break that reduces RCS pressure to the containment pressure?

b. What is the expected pressure in the accumulators and the CRTs after discharge and does this result in nitrogen being discharged into the RCS? If so, how is it accounted for in the W analysis?

c. How is the 60 psig external pressure specification used in the design of the accumulators and the core reflood tanks?

d. If the response to item b indicates nitrogen can be discharged into the RCS, what would be necessary on the part of the operators to isolate the accumulators and CRTs and how much time would be required to perform the operation af ter making the decision to do so?

RESPONSE

, a. The information was provided to the staff.

l l

l

b. Nitrogen is discharged f rom both the accumulators and the CRT's.

It is conservatively neglected in the EM model. An adiabatic l

assumption has been used for both the accumulators and the CRT s, but this assumption is being re-examined for the CRT's.

c. The internal pressure is the critical parameter in the design of these items.

l d. These operations are not permitted under design conditions.

O 440.172 (6.3-51) How is the minimum temperature that is listed in this table j 6.3 used with respect to the EWST?

O WAPWR-PSSS 440-110 AMENDMENT 1 2854e:1d MAY, 1986

~ RESPONSE:

- The value has no meaningful relationship to the design since it is not a critical item.

s 440.173 (6.3-52) The design heat transfer rate is given in this table as one 6.3 value and in Table 1.3-1 as another. Which is correct?

RESPONSE

The Chapter 6.3 value is correct. A corrected Table 1. 3-1 is O provided in Attachment 440.173.

440.174 What is the reason the RHR heat exchanger is designed for a tube side 6.3 temperature which differs from that of the low head pump miniflow heat exchanger tube side temperature, with the latter being consistent with the RHR pump design temperature?

RESPONSE

gs The RHR exchanger tube side temperature is based on the end of i

cooldown condition and the low head pump miniflow heat exchanger tube side temperature is based upon a dead headed pump.

440.175 Why is the tube side of the RHR heat exchanger designed for a flow 6.3 rate which dif fers f rom the RHR pump design flow rate for the same lineup configuration?

l i

! RESPONSE:

l One of the values includes miniflow.

440.176 The RHR initiation pressure is given on page 5.4-2 as approximately 6.3 400 psig; in Table 1.3-1 as approximately 425 psig. Which is correct?

RESPONSE

The 400 psig value is correct. The correction is provided in Attachment 440.173.

O WAPWR-PSSS 440-111 AMENDMENT 1 2854e:ld MAY, 1986

l 440.177 The minimum accumulator operating pressure is given in Table 1.3-1.

6.3 lable 6.3-2 lists the same value as the normal operating pressure.

Please clarify.

RESPONSE

The word " minimum" should be " normal". The correction is provided in Attachment 440.173.

440.178 The minimum core reflood tank operating pressure is given in Table 6.3 1. 3 -1. Table 6.3-2 has this pressure as the normal operating pressure. Please explain.

RESPONSE

See 440.177.

440.179 What is the meaning of the runout flow rate which is given in Table 6.3 1.3-1 as two dif ferent values?

RESPONSE

One is for conditions of spray alignment and the other for the RHR mode of operation.

440.180 Table 6.3-3, Integrated Safeguards System Relief Valve Data 6.3

a. Valve 8850 (high head pump suction f rom the EWST) is identified as having a specified relief setting. What is the behavior of this valve if the discharge side pressure is elevated? If the discharge side pressure is higher than the setpoint? If both the inlet and discharge sides are above the setpoint?

b. The setpoint of Valve 9022 (low head pump suction from the EWST) is specified in this table. The following statement is contained on page 5.4-2 in the last paragraph: "The RHRS is designed to be isolated from the RCS whenever the RCS pre.sure exceeds the normal RHRS cut in pressure. The RHRS is isolated f rom the RCS O

WAPWR-PSSS 440-112 AMENDMENT 1 2854e:ld MAY, 1986

on the suction side by two motor-operated valves in series on each suction line. Each of the normally closed motor-operated valves is interlocked to prevent its opening if RCS pressure is greater than approximately . . and to automatically close if RCS pressure exceeds ....? These pressures are in excess of the stated setpoint. Please discuss.

RESPONSE

a. The valve operates on the differential pressure setting.

b. This is a generic item. The staff is referred to RG 1.70.

440.181 Table 6.3-4, Failure Mode and Effects Analysis 6.3 NOTE: The staff will provide guidance on whether this table is necessary.

O a. Only single failures of the pump, selected valves, and a tempera-ture control channel are discussed. Please address the case of one diesel failing to provide power for operation of the ISS.

b. An item not covered in failure of 8807 to close is that the hot leg injection path could still be used, with injection into both the downcomer and the hot leg. The difficulty would be that there is no provision for the operator to determine how much flow is going by each path. However, if flow meter FI-920, which was in the preliminary submittal, was still in the system, such indication would be provided. Please comment.

c. Why is a number designation used for the pumps and a letter designation used for the valves rather than being consistent and using a letter designation for everything in the same subsystem?

l 440-113 AMENDMENT 1 l WAPWR-PSSS MAY, 1986 2854e:ld

l

d. In Item 4), temperature control channel T-866 is identified as having a failure mode high such that high injection fluid temperature is indicated, with the result that CCW flow is initiated to the shell side of the RHR heat exchanger before it is needed. Why is the reverse situation where it fails low not identified? The latter would appear to result in no flow to the shell side of the RHR heat exchanger, with the result that no cooling is provided to the train,

e. Portions of 440.71 are applicable to this table, specifically the first paragraph and item a.

f. There are a number of other valves associated with the flow path of the high head pump that are not mentioned in this table, apparently because they are not required to operate if they are in the " normal" position. What is the reason that mispositioning of these valves is not addressed along with the failure of the valve to respond to a signal to reposition to the correct position for SI operation?

RESPONSE

a. Two are lost for this single failure.

b. If 8807 cannot be closed, then the opening to the hot leg cannot be accomplished. However, there are still three more subsystems that can be used and this failure is therefore not of significance.

c. This is a historical method of doing things.

O ADDITIONAL NOTE: The remainder of this item was not discussed since a similar discussion took place in regard to Section 5.4.7 and it is not clear whether the table will be retained in the SAR.

O WAPWR-PSSS 440-114 AMENDMENT 1 2854e:ld MAY, 1986

440.182 Figure 6.3-1. What is the design pressure of the line between valves 6.3 9015 and 9016? The concern is whether the line will be overpres-surized if there is leakage in 9016 during operation of the high pressure SI system.

O RESPONSE:

This is designated as a 2501 line.

440.183 Please provide a discussion of check valve behavior with respect to 6.3 backpressure (or pressure differential) required to close, to open, and the conditions under which backflow can occur, if any.

RESPONSE

This is a generic item, and the SP/90 is no different from the other W plants.

440.184 (Figure 6.3-2 and attached notes) The staff does not understand a l 6.3 number of the pressure notations. Please clarify. For example, in the ECCS and containment spray modes:

a. Locations 9 and 11 are at a lower pressure than containment (23) and all of the lines that are connected to the EWST, which is at the same pressure as containment.

! b. The line leading f rom the accumulator is identified as at the accumulator set pressure downstream of the isolation valve, which is shown as open, with the RCS at a significantly lower pressure.

c. The line leading f rom the CRT is identified as at the CRT set pressure downstream of the isolation valve, which is shown as open, with the RCS at a significantly lower pressure.

d. Location 21 is downstream of locations 14 and 15, but its pressure is higher.

O WAPWR-PSSS 440-115 AMENDMENT 1 2854e:ld MAY, 1986 i

(

e. The temperature drop across the miniflow heat exchanger for the -l RHR pump is very small; much smaller than the staff would l anticipate for operation.  ;

And for the normal RHR mode:

O

f. The inlet temperature to the miniflow heat exchanger is shown as very close to the outlet temperature.

g. The pressure increase across the RHR pump appears to be slightly higher than the shut off head that is given in Figure 6.3-5.

h. The pressure difference through the RHR miniflow heat exchanger appears to be small for the flow rate that is given when compared to similar information for a lower flow rate on the prior page.

i. The flow rate or the prior page is given as the runout flow rate identified in Figure 6.3-5. However, the pressure dif ferential developed by the pump does not correspond to the pressure differential shown on the figure which corresponds to the runout condition.

j. The temperature at Location 22 does not appear to correspond to l any nearby temperature.

k. Significant pressure drops due to flow are shown in some portions of the figure, but none are shown in the suction line.

RESPONSE

O Table 6.3-2 and Figure 6.3-2 will be revised to reflect the latest ISS information as part of the Integrated PDA submittal. NRC's detailed comments will be addressed / corrected at that time. However O

WAPWR-PSSS 440-116 AMENDMENT 1 2854e:ld MAY, 1986

to facilitate your current PDA effort the following comments are provided:

a. The pressure at locations 9 and 11 should read psig (Note: (a,c)

RCS and containment pressure assumed to be equal).

_ (a,c)

, b. The pressure at node 19 should read _

the minimum accumulator pressure).

c. The pressure at node 18 should read _

the minimum CRT pressure).

d. The pressure at nodes 21 and 17 should be [ ] psig; it is(a,c) essentially the same as the pressure at nodes 15 and 14.

e. The small temperature drop across the RHR miniflow heat exchanger is correct.

O f. The expected temperature drop across the RHR miniflow heat exchanger is small.

g. The minimum developed head across the RHR pump is ft. during(*'CI RHR operation. This corresponds to a AP of { g psid - the (*ta,c)'C '

[ ]psid listed in nodes 13 and 14 is correct.

h. The pressure dif f'erence between nodes 15 and 16 will always equal the RHR/CS pump developed head.

O i. The minimum developed head of the RHR/CS spray pump is[ ]f t. at ta,c) runout flow. This corresponds to a aP of 7 psid - the

. psid listed in nodes 13 and 14 is correct.

j. The temperature at node 22 is the RHR HX outlet temperature and is the same as that listed for nodes 7, 8, 9, 10.

O WAPWR-PSSS 440-117 AMENDMENT 1 2854e:ld MAY,1986

k. The pressure drop in the pump suction lines is small and is more than compensated by the elevation head of the water (a.c) psig was shown to simplify unders'tanding, but (a.cl shouldread~_]psig.

[

440.185 Please identify all support services, such as direct cooling, lubrica-tion, operation air, and environmental temperature control, which are needed for the ISSS operation and which have not been previously identified. Also indicate if they are considered to be safety related or otherwise identify the reliability of those services being available under accident conditions.

RESPONSE

l The CCW dependencies are identified on the P& ids. In general, the valves are M0V's and are not required to move. This topic will be covered in the PRA and therefore further information will be presented with that documentation.

440.186 Please provide a discussion on leaks and breaks in ISSS equipment which communicates with the CCW system. Ir:clude a list of each ISSS component (or reference a list provided in response to another question). The concern is CCW system response and potential overpressure due to exceeding the capacity of relief valves. In the response, please cover common cause f ailures and the potential for affecting all plant CCW.

RESPONSE

See Module 13.

440.187 Please corrrnent on the large volume of injection water that can be pro-vided by the low pressure RHR pumps if they are aligned to inject into the RCS and whether the emergency operating procedures will contain steps for the necessary alignments, and the conditions under which these alignments would be used, if any.

O WAPWR-PSSS 440-118 AMENDMENT 1 2854e:ld MAY, 1986

RESPONSE

The capability is provided for this usage, and the RHR pumps can be held in reserve as a backup to the high head pumps. Emergency procedures should address this item. The staff is referred to Module 2 Section 3.1 item 7c.

440.188 Please address the impact of instrument uncertainty on Technical Specification limits and identify how safety analysis results are used to assure there is sufficient allowance for instrument uncertainty.

RESPONSE

Technical Specifications are not yet available. Instrumentation ,

uncertainty will be considered to obtain Technical Specifications from safety analysis set points.

440.189 What alternate ECCS injection locations were considered and why were these rejected in favor of the approach selected for the SP/907 At a minimum, please contrast injection into the lower plenum, upper plenum, upper head, downcomer, hot leg, and cold leg, including combinations of these locations.

RESPONSE

The only serious configuration considered was for loop injection of water. The SP/SO configuration provides for injection of water from all sources for intermediate and large breaks, whereas loop injection may have lost the water injected into one loop. Further, the SP/90 O configuration results in less cost due to the interactions of the the the component number of components needed and sizing of capacities.

440.190 (Page 15.5-2) "If a reactor trip does occur, the RCS pressure will 15.5 remain above the shutof f of the safety injection pumps so that no flow will be provided to the RCS." Please provide substantiation for this conclusion, including all applicable assumptions and suitable O

440-119 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

plots of RCS pressure and pressurizer level as a function of time.

Include a description of the sensitivity work and background experience utilized in your analyses. (See also 440.149a.)

RESPONSE

Reactor trip events are performed as part of system design transients. The severity of the cooldown transient following a reactor trip depends upon the assumptions made with regard to secondary system cooling. Based upon the results of the APWR reactor trip design transients, the RCS pressure can fall below the shutoff head of the safety injection system. Since the actuation of SI following a reactor trip is not a safety issue, the writeup on this section will be revised. The proposed amendment is shown in Attachment 440.190.

440.191 (Page 15.2-2, second paragraph) "This incident simply results in a 15.5 reactor trip." Please clarify the plant response with respect to spurious SI signal with respect to operation of the charging pumps insofar as they are injecting water into the RCP seals. (See also 440.149b.) Include a discussion of operator actions and the process whereby the operator will determine the reason for the SIS actuation. (A suitable reference may be used for the operator

actions.) Provide a list of all equipment which will be initiated or isolated by a spurious SI signal.

RESPONSE

With offsite power available, the charging pumps will continue to run following an "S" signal so that RCP seal integrity is maintained.

Module 13 " Auxiliary Systems" will clarify the CVCS design. The charging pumps do not stop on an "S" signal unless there is also a i

loss of of fsite power. In this event, seal cooling is provided by the backup seal injection pump and by component cooling water.

i i

l WAPWR-PSSS 440-120 AMENDMENT 1 2854e:ld MAY, 1986

Although specific operator actions have not been formulated at this time, similar operator actions to those for standard plants would be followed.

The safety injection signal will actuate the following engineered safety features:

1. Startup of emergency feedwater pumps, steam generator letdown isolation, and startup feedwater termination;

2. Startup of emergency diesels;

3. Start of service-water pumps and isolation of non-essential service water, if required;

4. Start of other pumps (e.g., component cooling);

5. Start of emergency fan coolers;

6. Start of high head safety injection pumps;

7. Safety injection diesel loading sequence when and if sequencing is necessary;

8. Reactor trip, provided one has not been generated by one of the reactor trip functions identified in Section 7.2; O 9. Phase-A containment isolation to prevent fission product release; i.e., isolation of all lines not essential to safety injection;

10. Containment ventilation isolation;

11. Control room isolation;

{

12. Feedwater isolation.

440-121 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d t

The seal injection line will have a special f ail closed, air operated globe valve that will serve as the containment isolation valve, outside containment. This valve will receive a "T" signal to close.

However, the spring that closes the valve will only be able to resist about 80 psi from the charging pumps before it would be forced open.

l As a result, any of the two charging pumps or the backup seal injection pump can provide seal injection following an "S" signal.

Independent seal cooling is also provided by the component cooling water (CCW) which is not isolated on an "S" signal.

Phase B containment isolation is not actuated by an "S" signal.

Phase B isolation does not affect RCP seal injection or CCW cooling to the thermal barrier.

i HVAC actions are covered in Module 13 " Auxiliary Systems".

440.192 (15.6-1A, Section 15.6.2.1) l 15.6

a. "The estimated f requency of a primary sample or instrument line  :

1 rupture classifies it as a limiting fault incident." What is the estimated frequency of primary sample or instrument line rupture? Provide the documentation for this conclusion. Are there any lines which lead outside of containment f rom the RCS pressure boundary that are not included in the estimated frequency value? If so, what lines and, what is the reason for their exclusion?

b. Please provide information describing all lines f rom the primay O, system penetrating containment. Include location (angular location, elevation), diameter, a description of the penetration (such as a drawing), identification of valves, and isolation actions. Include the information with respect to both the ..

primary and secondary containment buildings (or structures).

O WAPWR-PSSS 440-122 AMENDMENT 1 2854e:ld MAY,1986

RESPONSE

a. The statement that "The estimated f requency of a primary sample or instrument line rupture classifies it as a limiting fault incident," is not correct. It is considered a Condition 11 event ta,d l

of moderate f requency. l O _

For the 3/4-inch instrument or sample lines connected to the pressurizer vapor space, no restrictor is required since the mass flow from the vapor space can be matched by a single charging pump. The failure of a vapor space piping connection would however result in a decrease in system pressure because of limited steam generation capability in the pressurizer and would eventua.11y result in safety system actuation.

l All of the lines which penetrate containment have the capability to be remotely isolated by the operator. Since the letdown line is the largest open line from the primary system which penetrates containment, the rupture of this line is considered to be the limiting event of this type.

The basis and frequency of ruptures in the instrument and sample lines are documented and considered in the plant transient initiating event f requency provided and discussed in Module 16, O "Probabilistic Safety Study", Tables 1.1 -2, 1.1 -4 a nd 1. 2-3.

b. Detailed information on the angular location, elevation, and penetration detail for all lines from the primary system

% penetrating containment is not available at this time. However, l

a description of the piping penetrations was provided in respon',e to question 440.23 on Module 3, " Introduction and Site". Of these, the only additional connections directly from the primary 440-123 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY, 1986 i

.-r,. , - - - - . - - . , . . - , - . . . - , . - - . - . . - . - - - - - -- , - , - . -

l system penetrating containment are the four RHR letdown lines.

These 8-inch lines are isolated by two closed, power removed, interlocked and alarmed gate valves in series in each line. The only other penetrations are the CVCS letdown lines and instrument and sampling lines discussed above in part a).

440.193 (15.6-l A, Section 15.6.2.1) "Following such a break, the flow out of 15.6 the break would be limited by the letdown orifices to approximately 100 gpm, assuming that the orifices were selected for normal letdown."

a. Why was this assumption made as contrasted to the larger letdown rate that is possible if the bypass is used?

b. What are the implications with respect to the limiting fault accident if the higher flow rate is used?

c. What is the likelihood that the higher rate letdown will be used to enhance emergency boration?

d. Are there any lines f rom the reactor coolant pressure boundary which lead outside of containment that are larger than the letdown orifice (with respect to possible leak rate) and, if so, why were they excluded?

e. Discuss the conservatism of the flow rate taken for tha letdown line as contrasted to the staff review approach as outlined in the SRP on page 15.6.2-3, item c, which states "The amount of primary coolant released is conservatively estimated by assuming critical flow at the small line break location with the reactor coolant fluid enthalpy corresponding to normal coolant operating conditions. The reviewer evaluates the reactor coolant release rates provided by the applicant, taking into consideration similar information for plants recently reviewed. The reviewer should verify the release rates and the total amount of coolant released with the RSB in a coordinating review effort."

WAPWR-PSSS 440-124 AMENDMENT 1 2854e:ld MAY, 1986

RESPONSE

a&b The CVCS design that is currently invisioned for Module 13 has only one letdown orifice inside containment and it is rated for maximum letdown, or gpm. The CVCS will also have a control

valve outside containment that will be used to reduce the letdown

\ flow to values less than gpm as desired by the operators; f or

example to normal letdown flow of _

gpm. As a result the (* *CI maximum flow from a letdown line break outside containment would

~~

be gpm. This is an isolatable break, in fact low pressurizer (**c)

~

leve1 will automatically close redundant valves in the letdown line which will stop the loss of RC. The flow meter in the letdown line is not necessary to protect against this event.

c. Safety grade boration will not utilize the CVCS letdown equipment but instead will use separate emergency letdown lines that connect the RCS hot legs directly to the EWST. The emergency letdown is sparged into the EWST so that subcooling via heat exchangers is not required. These emergency letdown lines are designed for[]gpm letdown directly f rom the RCS HL and are (a,c) continuously sloped downward to the EWST and are well supported.

This modification will be shown in the integrated PDA.

4

d. The CVCS letdown line is the largest normally open line that is i

connected to the RCS that also penetrates the containment. There are larger lines in the ISS but they all have either multiple check valves that are tested every startup or as in the case of the RHR letdown line are isolated by redundant MOV with diverse interlocks and power removed.

l

e. As mentioned in part a. above the maximum CVCS break flow occurs with subcooled flow (i .e. with the letdown HX operating). The maximum flow rate of gpm is calculated with minimum pipe, (8.cl valve, HX, and orifice resistances.

O 440-125 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

"The operator would receive a low flow alarm from the 440.194 (Page 15.6-18) 15.6 letdown flowmeter, and would be required to isolate the break manually." The P&ID information provided via a supplementary submittal (June 1984) as a part of Module 4 (Reactor Coolant System Piping end Instrumentation Diagram) is illegible. Hence, the staff is unable to evaluate the location of the flowmeter and the isolation valves as contrasted with possible break locations. Please provide this information and, in addition, provide legible copies of the P& ids.

O

RESPONSE

The flow meter is not required to protect against letdown line breaks. A complete break of the letdown line is discussed in the answer to 440.193. A partial break of the letdown line would not necessarily cause a low pressurizer level because of the capability of makeup to the CVCS/RCS. In this case the protection is provided by building area radiation monitors. In both cases the letdown flow meter and the makeup to the CVCS volume control tank may also alert the operator but is not relied upon.

440.195 (15.6-18, last paragraph) The acceptance criteria of 10CFR50.46 are 15.6 paraphrased in a list, and the wording is different when compared to 10CFR50.46. This was discussed with W personnel on September 26, 1983, and the staff understood the wording would be changed to be l

consistent with that stated in 10CFR50.46. (See items 68, 69, and 70 of the list pertinent to review of the preliminary Module 1 dated June 1983.) The wording in the version being reviewed (October, 1983) has been changed, but still appears inconsistent with the rule. Item b originally stated: "The amount of fuel element cladding that reacts chemically with water or steam does not exceed one I percent of the total amaunt of Zircaloy in the reactor." This would have permitted more hydrogen to be generated than permitted in the rule, and hence would have permitted more chemical reaction. The statement now is "The amount of fuel element cladding that reacts

{ chemically with water or steam does not exceed one percent of the O

WAPWR-PSSS 440-126 AMENDMENT 1 2854e:ld MAY , 1986 l

total amount of Zircaloy in the fuel." There is no Zircaloy in the fuel; hence no hydrogen generation is permitted. This is acceptable, but is it in accord with M's intent?

O

RESPONSE

Westinghouse can clarify the statement further. See our revision in Attachment 440.195. It is our interpretation that the fuel rod contains both the fuel pellets and cladding.

440.196 (15.6-2, item a) Previously submitted SARs for M plants, in this

. 15.6 section of the report, typically contain the words "However, no credit is taken in the LOCA analysis for boron content in the injection water. In addition, the insertion of control rods to shut down the reactor is neglected in the large break analysis." These words are missing f rom the SP/90 report. Please clarify the rod assumptions for the Ch.15 analyses, contrast the approach to that l applied to prior M plants, and justify any changes. (See also 440.211, 440.222 and 440.223.)

RESPONSE

Assumption is identical to that made for prior Westinghouse plant analyses: No credit is taken for control rods or boron content of the injection water.

440.197 Please discuss applicability of the previously approved LOCA codes to 15.6 the SP/90 plant with respect to differences regarding such items as:

a. Upper vessel elevation and configuration

b. Downcomer length O c. Bypass during blowdown

d. Injection location and nozzle configuration O

440-127 AMENDMENT 1 MAPWR-PSSS 2854e:1d MAY,1986

e. Hot wall effect

f. CRTs

g. Fuel configuration

h. Bypass between the upper vessel and the downcomer upper annulus

1. Core former/ reflector region

j. Hot and cold leg elevations with respect to other portions of the RCS See also 440.222 and 440.223.

RESPONSE

Westinghouse has looked in great depth at the LOCA transient results and unique APWR design features with respect to the applicability of previously approved LOCA codes and models. As a result of this review, Westinghouse has put together a LOCA ECCS evaluation model development program which includes sensitivity studies with current evaluation model codes, evaluation model prediction comparisons to more detailed CO6dA/ TRAC calculations, a LOCA test program, and use of new computer codes for the reflood portion of the transient.

l Our position on each of the items is discussed below, l

a. Upper vessel evaluation and configuration.

Probably the biggest uncertainty in our current evaluation model calculations is the ability to handle the complex upper internals geometry. Because of this, Westinghouse has performed upper plenum noding sensitivity studies, performed C/T analyses to further investigate the complex 3-0 hydraulic behavior, and developed a test program to confirm two phase pressure drop and liquid accumulation during reflood.

WAPWR-PSSS 440-128 AMENDMENT 1 2854e:ld MAY, 1986

e The LOCA tests are expected to confirm the C/T pressure drop predictions and will ultimately show the evaluation model predictions are conscrvative. Testing began in the fall of 1985 and will continue until the end of 1986.

The upper plenum metal heat release is calculated as described using the methodology described in WCAP-8302. It is standard Westinghouse methodology to lump all of the metal mass and surface area together for a given region. The overall heat transfer is kept constant and is input into the code. This value was determined to be conservative for Westinghouse standard plant LOCA transients in WCAP-8341. Because the APWR rod guides have a lot of surface area compared to the metal mass, they tend to release energy quicker than thick metal masses. A sensitivity study was performed with the heat transfer coefficient doubled.

The energy release increased and the effect on PCT was small.

Furthermore, C/T results, which model each metal individually and calculate time dependent heat transfer coefficients, show that the SATAN model heat release was greater,

b. Downcomer length The downcomer length provides the driving head for reflooding the core. Concern related to the downcomer length include the hot wall delay time and downtomer voiding due to metal heat input during reflood period. The hot wall / delay time is calculated using the CREARE hot wall correlation. The metal heat release to the downcomer fluid is calculated based on the wall temperature gradient and heat transfer coefficients based on fluid conditions. This downcomer model is the same model used in standard Westinghouse plant analyses using the BASH code.

O 440-129 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:ld

9

c. Bypass during blowdown Bypass during blowdown is handled similarly to prior Westinghouse plant analyses. ECCS water injected prior to end of bypass is subtracted from the vessel inventory at the beginning of refill.

d. Injection location and nozzle configuration Safety injection is handled similarly to standard plant analysis. One dif ference is injection location. In blowdown, this is handled conservatively since no downward momentum is considered for penetration into the downcomer to provide a benefit for the end of bypass calculation. It is believed that this benefit outweighs the potential negative effect on the negative core flow period. This is not considered to have a large effect on the transient based on the fact that the safety injection and core reflood tanks start to deliver near the end of blowdown and the contribution of the fluid f rom these nozzles is small compared to the flow penetrating the downcomer f rom the accumulators. However, no analyses have been performed to support this claim.

In reflood, the direct injection nozzle location poses an area of concern af ter the accumulators empty. High velocity steam f rom I the intact loop will enter the vessel and can entrain liquid from the downcomer thereby decreasing downcomer head. This is modeled using the Steen-Wallis correlation in the BASH computer code.

This correlation is considered to be conservative for this application because of the geometry of the flow path. Fluid flowing across the downcomer would contact the vessel wall, core barrel, and hot leg nozzle stubs, which would tend to separate the water from the steam.

O WAPWR-PSSS 440-130 AMENDMENT 1 2854e:ld MAY, 1986

e. Hot wall effect U.S. NRC mandates the use of the CREARE hot wall delay correlation reported in the technical note TN-202. The tests were 1/15 scale based on PWR type geometry.

The variables which strongly af fect the hot wall delay period were identified to be inlet water temperature, initial wall temperature, downcomer gap spacing, and downcomer length. The correlation was found to be capable of predicting the hot wall delay time as these variables were varied over a wide range.

The APWR dif fers f rom Standard PWR's in downcomer gap spacing and downcomer length; inlet water temperature and wall temperatures

, are consistent. Tests were performed varying the L/S ratio from l 9-72. Since the downcomer length to gap ratio for the APWR is about 20, its downcomer falls well within the range of the test data.

Current WREFLOOD calculational results show that the hot wall delay for the APWR is approximately .3 seconds which compares to approximately .05 seconds for the standard 412 plants.

f. CRTs The core reflood tanks are currently modeled similarly to accumulators using the adiabatic expansion model (PVy =

O- constant). This assumption will be evaluated in the future when the reflood sensitivity studies are completed. It poses no problem at this stage of the design since the CRT line resistance can be adjusted to give the required flow.

i I

O WAPWR-PSSS 440-131 AMENDMENT 1 2054e:1d MAY, 1986

l

g. Fuel configuration 1

The APWR fuel assembly configuration is a 20 x 20 fuel rod lattice with 16 2 x 2 guide thimbles and/ instrumentation thimble with an active fuel length of 12.792 feet. The concerns of the new fuel design configuration pose on an evaluation model calculation are: 1) effect of large thimbles on core heat transfer, 2) the handling of the significant water volume inside the thimbles, 3) effect of rod size and length on FLECHT heat transfer. The assessment of the large thimble effect on blowdown heat transfer can be made by evaluating the Combustion Engi-neering film boiling data on 5 x 5 rod bundle tests with and without a large thimble. For reflood heat transfer, FLECHT test data shows the placement of a thimble or failed (unpowered) fuel rod simulator overall improves the heat transfer adjacent power rods because the thimble or unpowered rod acted as a radiant energy heat sink. The effect of larger thimble tubes is to serve as larger heat sinks over existing Westinghouse fuel designs due to the larger surface area. The existence of thimble tubes also results in different power generation-flow area ratios for flow channels with and without a thimble tube. The dif ferent steam mass generation rates cause steam flow among channels. This interim impacts droplet movement and results in droplet redistribution among the channels. This effect can have a I

potential negative effect on reflood heat transfer. A hand i

calculation was performed conservatively assuming migration of the droplets to the cooler walls of the thimble. This phenomena was calculated to be at most a 70'F penalty on reflood clad temperature. No credit was taken for the radiant heat sink capability of the thimbles in this study. Based on the FLECHT test data and if credit is taken for the radiant heat sink capability of the thimbles that the potential reflood heat transfer penalty is small. There is, at this time, sufficient margin between the blowdown and reflood clad temperature peaks to accommodate a small reflood penalty if necessary.

O 440-132 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY, 1986

a The assessment on the impact of the water volume and flow area of the thimbles was done by performing sensitivity studies explicitly modeling the thimbles. These studies show the impact of the thimbles to be minimal.

O Reflood heat transfer is based on the FLECHT correlation developed from test data on past Westinghouse fuel designs (15 x 15 and 17 x 17). APWR fuel rod geometry parameters lie within the range of these tests such that the FLECHT correlation can be applied conservatively to the APWR analysis. Also, the skewed power methodology used for cores with lengths greater than 12 feet (such as used for the South Texas 414 design) was used for the APWR analysis.

h. Bypass between upper vessel and downcomer upper annulus.

This is handled as in typical Westinghouse plants with the upper i head bypass flow modeled. This bypass flow path is modeled similarly to other flow paths in the SATAN model with the appropriate flow area, hydraulic diameter, loss coefficient, etc. For most of the blowdown transient, the fluid flows f rom the upper head to the downcomer annulus.

1. Core former/ reflector region The radial reflector region is not modeled in the APWR standard analyses. Blowdown sensitivity studies have been performed with O the evaluation model where the radial reflector has been explicitly modeled. The results show the ef fects to be small.

The more detailed COBRA / TRAC analysis results support this conclusion. The impact of the radial reflector. energy release O during reflood was evaluated and conservatively shown to have only a small impact on flooding rate (5-10 percent) and approximately 20*F on reflood clad temperature using the WREFLOOD code. This unconservatism does not impact PCT since it occurs O during blowdown.

440-133 AMEN 0 MENT 1 WAPWR-PSSS MAY, 1986 2854e:1d l

l

j. Hot and cold leg elevations with respect to other portions of the RCS. .

Elevation changes are incorporated into the analysis as in standard Westinghouse plant analyses. Pressure differences include a term for elevation changes. Flow paths that have large elevation changes can impact drift flux correlation predictions.

Sensitivity studies have been performed such as the downcomer noding study (where the number of downcomer nodes was doubled) reducing the flow path elevation difference. The results show a very small impact on the hot channel calculated clad temperature transient. Flow paths with large elevation changes tend to resemble homogeneous behavior.

440.198 (15.6-4, first paragraph) "The high head safety injection pumps will 15.6 inject borated water whenever the reactor coolant system pressure is below [ ] psia." Figure 15.6.4-38 shows the shut-off head to be 300 psi lower than that. Please clarify.

RESPONSE

Nominal shutoff head for the safety injection pumps is

[ ]". The curves in Figure 15.6.4-38 are used for conservatism. For the large break analyses presented only the lower portions of the curve is important since SI does not deliver until the pressure is less than 200 psi.

440.199 (15.6-4, first paragraph) "Since the loss of offsite power is 15.6 assumed, the reactor coolant pumps are assumed to trip at the inception of the accident." Please justify this assumption. Include establishment that the assumptions made regarding RCP operation for the transients and accidents being analyzed are conservative with respect to the expected RCP operation which will result from resolution of Generic Letter 83-10C. (See also 440.223.)

O 440-134 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY, 1986

e

RESPONSE

Sensitivity studies were performed for large break LOCA transients with and without offsite power available. Results show case with

' reactor coolant pumps running shows blowdown PCT benefit due to the improved positive core flow period.

This assumption is consistent with the Westinghouse position in Section 6.4 item 92 of Module 2, " Regulatory Conformance".

O 440.200 (15.6-4, second paragraph) Page 6.3-30 contains the statement "the 15.6 time at which end of blowdown occurs is determined by zero break flow, which is a result of achieving pressure equilibrium between the RCS and the containment." Page 15.6-4 has the statement "The blowdown phase of the transient ends when the RCS pressure (initially assumed at 2250 psia) falls to a value approaching that of the containment atmosphere." Please clarify. (See also 440.197c.)

RESPONSE

In the Westinghouse ECCS evaluation model, end of blowdown is calculated when any of the following conditions are met: 1) downflow in the downcomer exceeds 100% of ECCS flow, 2) break flow on the vessel side is zero, or 3) the lower plenum is filled with liquid.

All three of these coditions can occur as the RCS pressure falls to a value approaching the containment atmosphere. In most cases, condition (2) above is the first to be satisfied.

These phrases in pages 6.3-30 and 15.6-4 are essentially correct in the manner they are used.

l 440.201 (15.6-4, third paragraph) "The reflood phase of the transient is 15.6 defined as the time period lasting from the end of refill until the reactor vessel has filled with water to the extent that the core temperature rise has been terminated." How is this definition consistent with a situation where the temperature is still high (but no longer increasing) and where hydrogen is being produced due to O

WAPWR-PSSS 440-135 AMENDMENT 1 2854e:1d MAY, 1986

chemical reaction of the zirconium with water and steam? How does it assure no further clad degradation? How does it assure that a coolable geometry results as required in 10CFR50.46(b)(4)?

RESPONSE

For the APWR, the hypothetical situation has not been seen in any calculation to date. The definition of the reflood phase remains valid.

440.202 (15.6-5, first paragraph) "Approximately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> af ter initiation of 15.6 the LOCA, the ECCS is realigned to supply water to the RCS hot legs in order to eliminate the potential for a high concentration of boric acid in the reactor vessel." Is this always necessary for the SP/90 design? For example, if the "incore" thermocouples continuously show subcooling and one can thereby establish that there is no boiling in the core, would it be necessary? The concern is elimination of unnecessary operations which could perturb satisfactory core cooling. (Note the closely related 440.135.)

l RESPONSE:

j The four high head pumps are normally aligned to deliver coolant to the reactor vessel downcomer via four separate direct vessel injection (DV1) nozzles, in the event of a hot leg LOCA, subcooled flow would be established through the core, without any operator action, as soon as the core decay heat decreased to a value that could be removed by the coolant injected into the downtomer. For this accident no realignment would be necessary, provided that the operation has sufficient diagnostic information to determine the thermal hydraulic state of fluid in the core.

However, in the event of a cold leg LOCA, subcooled flow could not be g established through the core if the high head pumps remain aligned to W the DVI nozzles. For this event, the operating high head pumps must be temporarily aligned to the RCS hot legs in order to back flush the l

buildup of high concentration of boric acid from the reactor vessel.

WAPWR-PSSS 440-136 AMENDMENT 1 2854e:ld MAY,1986 l

For the worst case scenario, it must be assumed that a loss of offsite power and a failure of one of the two emergency diesel i generators could occur simultaneously with the postulated cold leg LOCA. Therefore, only two of the four high head pumps would be operating and delivering to two of the DVI nozzles. For this case, it is recommended that the two operating high head pumps be realigned to deliver to the corresponding two RCS hot legs to expedite the back flush of the high boric acid concentration from the core.

O Af ter a short period of time with only hot leg injection, it is recommended that one of the two operating high head pumps be realigned to the DVI nozzle so that a simultaneous delivery of coolant to both hot leg and DVI is established for the long term core cooling phase.

440.203 (15.6-5, second paragraph) This discussion breaks the small break into three stages: gradual blowdown, core recovery, and 15.6 LOCA long-term recirculation. What is the meaning of " core recovery" since W claims the core never uncovers?

RESPONSE

The term " core recovery" is meant to describe the time period af ter the core mixture (level within vessel barrel) reaches its minimum value and is increasing. For the APWR, a better term would be

" vessel level recovery".

440.204 (15.6-6, first paragraph) Reference is made to the SATAN-VI,

\ 15.6 WREFLOOD, C0CO, and LOCTA-IV codes, which are discussed as described l

in detail in Ref erences 15.6.4-4 through 15.6.4-7. Recent SARs on W plants have identical wording, but reference is also made to modifications to those codes in more recent documentation. Reference

\ to the modifications is missing in the SP/90 discussion. Please clarify with respect to the codes actually used, whether they are the most recent W code versions, including application techniques, and O

440-137 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:ld

whether they are approved by the NRC in the version used for the SP/90 application.

RESPONSE

The codes used in the SP/90 analysis are part of the 1981 version of the Westinghouse ECCS evaluation model approved for use by the NRC l

(see page 15.6-8) .

l 440.205 (15.6-7, third paragraph) "An estimated containment pressure transi-15.6 ent was input into the thermal-hydraulic codes due to insufficient

data available for containment design." What are the plans relative to investigation of containment response? Will the estimated input be compared to actual calculated response? If so, where will it be reported?

RESPONSE

Specific APWR containment pressure analyses have not been completed to date. They will be reported to the staf f upon completion. The final reflood analysis, using the BASH code, will use the APWR specific backpressure transient.

l 440.206 (15.6-7, first paragraph) "During the reflood portion of the transi-15.6 ent, the core is assumed to remain in a subcritical condition due to the boron content of the injection water." Provide substantiation for this assumption, l

RESPONSE

(**CI Nuclear fuel design calculates a critical boron coacentration of ppm under cold all rods out condition to_ maintain 1% subcriticality (a,c) Since concentration of ECCS water is ppm subcriticality in excess of 1% is assured.

440.207 (15.6-8, second paragraph) "Also, the analysis in this section was 15.6 performed with the upper head fluid temperature equal to' the RCS cold leg fluid temperature, which is achieved by passing a sufficient WAPWR-PSSS 440-138 AMENDMENT 1 2854e:ld MAY, 1986

4 .

amount of bypass flow into the upper head." Is this limited to the

' initial condition prior to initiation of the break or is there some other meaning? What is the bypass flow rate and how was it  !

determined? (See also 440.213.)

O' RESPONSE:

Bypass flowrate of total reactor flow is calculated to

maintain upper head at the RCS cold leg temperathre. This flow is I determined by the reactor equipment system design group using 3-D analyses to calculate the bypass flow pressure drop and then designing the spray nozzles with the appropriate flow area such that there is sufficient flow to keep the upper head at the cold leg _

! temperature under steady-state conditions. The bypass flow and (a'c)

I upper head temperatures are used as initial conditions prior to the 1 initiation of the break.

440.208 (15.6-8) The description of the Small Break LOCA Evaluation Model 15.6 contained on page 15.6-8 does not mention liquid holdup in the upper vessel or steam generator tubes during any portion of the transient.

Please address this subject with respect to the phenomenon and the ability of the codes to properly model the behavior. Include a discussion of flow between the upper vessel and the upper downcomer, l and between the pump discharge and the pressurizer due to the pressurizer spray lines.

RESPONSE

in order to simulate liquid holdup, the homogeneous mixture option of O WFLASH is used in the uphill side of the steam generator tubes (Nodes 3 & 5 of Figure 440.2188). When steam flows into the tubes from the reactor vessel with the uncovering of the vessel nozzles, a two-phase mixture will exist in the tubes. The trapped liquid in the tube O mixture offers resistance to steam relief, leading to additional conservatism in the analysis. Further, any benefit from the possible draining back of the tube liquid into the reactor vessel when the steam velocity becomes sufficiently low (flooding phenomenon) is O conservatively ignored.

440-139 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY , 1986

Because the uphill sides of the tubes are homogeneous nodes, mixture level in these nodes becomes meaningless. In a homogeneous node, the i mixture level is always at the top of the node. To quantify the transient liquid content in the uphill side of the tubes, quality versus time is shown in Figures 440.208-A through 440.208-C. These figures show the quality in the uphill side of the tubes for break sizes of 3 inch, 4.313 inch and 6 inch diameters. When the loop seals uncover, liquid is still being held in the uphill side of the tubes. At the time the loop seals uncover, the magnitude of the quality is:

Quality In Uphill Break Size Side of the Tubes

- - t a ,t) 3" 4.313" 6"

The flow path from the upper downcomer, thro gh e upper head to the vessel upper plenum is simulated by WFLASH. During a small LOCA fluid in the upper head initially drains into the upper plenum and the upper downcomer, af ter the upper head drains to a point where the static head of the remaining liquid is less than the pressure in the core, steam from the core is then forced backwards thru the upper head to the downcomer and out of the break. This flow path then provides a small vent path between the core and the break.

The spray line between the pump discharge and the pressurizer is not simrlated as a vent path in the WFLASH model.

l 440.209 (15.6-9, first paragraph) A reference is made to Figure 15.6.4-38, 15.6 which shows the safety injection flow rates assumed for the one and two high head pump combinations as a function of pressure. A comparison of this figure for the one pump condition and the Table 6.3-2 (page 6.3-48) information shows that the figure has a pump shut WAPWR-PSSS 440-140 AMENDMENT 1 2854e:ld MAY, 1986

i off head of about 280 psi less than the table. The figure runout flow rate is about 14 lbs/sec less than the table. Are these two sources of information consistent when one takes into account the resistance of the SI lines and the pump miniflow bypass behavior?

Part of the difficulty is that Figure 6.3-4 already has the shutoff flow rate incorporated into the shutoff head, and a pressure differential of 280 psi appears inconsistent with no flow through the SI lines.

4

?

RESPONSE

Figure 6.3-4 depicts the "as manufactured" minimum pump performance curve and allowable tolerances, while Figure 15.6.4-38 shows the injection flow from one (and two) HHSI pump (s) which is used in ECCS safety analyses. This safety analysis HHSI injection flow curve is established considering pump miniflow and injection piping resistance and also uses a conservative pump curve, i.e., the pump head / flow is lowered to provide conservative injection flows. Thus, the injection O flows vs. RCS pressure shown in Figure 15.6.4-38 will and should be conservatively lower than the flows obtained directly from the pump "as manufactured" head / flow curve.

i l 440.210 (15.6-9, first paragraph) "It should be noted that analysis has also 15.6 been performed for a 6 inch break with only one high head pump delivering coolant to the reactor vessel, and the results of that i analysis show no core uncovery with no credit for accumulator flow."

Please provide pertinent information concerning this calculation.

Include CRT behavior.

RESPONSE

Attached are Figures 440.210-A thru 440.210-G which show the results of a 6 inch break with only one high head pump operating.

O Accumulators and CRTs were not assumed to function for this case.

O 440-141 AMENDMENT 1 WAPW'-PSSS MAY,1986 2854e:1d l

440.211 Please provide direct vessel injection nozzle detail, including the -

15.6 vessel wall penetration and nozzle within the vessel that are pertinent to flow area and flow directionality. See also 440.197d.

RESPONSE

Direct vessel injection nozzle drawings are attached.

440.212 (15.6-9, third paragraph) "This is limiting for the small break 15.6 analysis because of the core uncovery process for small breaks. If the core uncovers, the cladding in the upper elevation of the core heats up and is sensitive to the local power at that elevation. The cladding temperatures in the lower elevation of the core, below the two phase mixture height, remain low. The peak clad temperature occurs above 10 ft." How is this consistent with no uncovery during a small break. Does the distribution matter?

RESPONSE

The axial power distribution is not important if the core remains covered.

440.213 (15.6-10, second paragraph) "The bases used to select the numerical 15.6 values that are input parameters to the analysis have been determined from extensive sensitivity studies on Westinghouse four loop N:>SS designs .... " .hese bases were developed roughly ten years ago.

l Please provide justification that they apply to the significantly dif ferent design of the SP/90. Address differences in respect to elevation, flow path length, nodalization, volume configuration and size, injection configuration, treatment and modeling of metal mass in tne upper vessel regions, and the impact of this structure on h

reflood behavior. See also 440.197 and 440.223.

RESPONSE

At the time the SP/90 analyses described were performed, the input parameters, noding scheme, assumptions, etc. were based on standard analysis input determined f rom past sensitivity studies. Since that O

440-142 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

d time, SP/90 plant specific sensitivity studies have been performed to support the bases for selection of meerical values that are input parameters. However, some previous sensitivity study results are still deemed applicable to the SP/90 design and were not repeated.

See also Westinghouse response to 440.197.

O 440.214 (15.6-10, third paragraph) " Based on the results of the past LOCA 15.6 sensitivity studies ..., the limiting large break was found to be the double-ended cold leg guillotine (DECLG). Therefore, only the DECLG

! O break is considered in the large break ECCS performance analysis."

Please justify this conclusion for the different SP/90 geometry and l elevations. Address the contribution of the CRTs and the applica-bility of the conclusion with this design innovation. Include consideration of the significant cooling that occurs due to downflow during the large cold leg break and contrast this to the behavior one j would expect for a large hot leg break. Address the behavior due to nitrogen from the CRTs. See also 440.197 and 440.223.

O RESPONSE:

Again, at the time the SP/90 plant analysis was performed limiting assumptions from past LOCA sensitivity studies were used. Since that time, Westinghouse has evaluated a spectrum of break locations, types, and sizes. Comparison of the DECLG break and the DEHLG break l shows a higher blowdown PCT for the DECLG due to the poor positive core flow period and a better blowdown cooling effect due to downflow of upper plenum and upper head water. The DEHLG always has a positive core flow.

The core reflood tanks contribution is minimal to the blowdown j calculation. Their contribution to the reflood transient should be fairly the same for the two breaks.

The effect of nitrogen addition to the system would be a reflood benefit previously shown for past Westinghouse designs. Since the O

WAPWR-PSSS 440-143 AMENDMENT 1 2854e:ld MAY, 1986

downcomer is full when the nitrogen is introduced, it would tend to -

stay in the upper downcomer annulus due to density differences between the water and the nitrogen gas (also see response to 440.226).

440.215 (15.6-11) This page lists the parameters which describe large break 15.6 ECCS results. Please also provide downtomer flow rate at a position removed from the entrance regions, CRT flowrate for all times during which fluid is moving f rom the CRT, and show treatment of water that was injected prior to the end of bypass, including the quantities.

For the limiting break analyzed, please provide pressure in the downcomer annulus, upper plenum pressure, and liquid carryover rate during reflood; and discuss the disposition of liquid.

RESPONSE

Plots of downcomer flow, CRT flow, downcomer pressure, upper plenum pressure, and liquid carryover rate during reflood are provided in the .8DECLG case in Figures 440.215 ( A thru H) attached. CRT flow during reflood is shown in Figure 15.6.4-18 in Module 1.

O Treatment of water injected prior to end of bypass is consistent with the accumulator bypass model described in WCAP-8302. Prior to end of bypass 99264 lbm of ECCS water was injected into the system, of that 49518 lbm was calculated to exit through break. The remaining 49746 lbm is called bypass deficit and is subtracted from the vessel inventory at the time of the switchover from SATAN to WREFLOOD.

Concerning the disposition of liquid carryover in reflood, the WREFLOOD model is a simplistic model and does not allow for mass storage. The liquid f rom core is assumed to flow to hot leg when it l 1s superheated in the steam generator. More detailed analyses of the liquid mass storage in the upper internals is currently being performed using the BASH code. The BASH analyses show that the core reflooding transient is improved when the liquid mass storage in the upper plenum is maximized by modeling techniques. Additionally, LOCA tests are scheduled to be perforned to evaluate the impact of the SP/90 upper internals on reflood behavior.

440-144 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:ld

440.216 Please discuss fluid behavior in regions between the core (core 15.6 former-reflector region) and the downcomer wall during LOCAs.

Include impact on vessel structure, if any, and on fluid behavior in the remainder of the RCS, if any. See also 440.197.

O RESPONSE:

The current evaluation model sensitivity study case with the radial

' reflector showed the fluid to flash to steam due to the rapid depressurization and heat release f rom the metal. Its effect on the transient is minimal. Preliminary COBRA / TRAC results confirm this behavior. LOCA forces analyses (subcooled loads) show a peak pressure differential of 26 psi between the core and the radial I

reflector. SATAN sensitivity studies show pressure difference to be much less during blowdown transients.

440.217 Please discuss LOCA behavior as a function of a locked rotor in an RCP 15.6 vs. a f ree rotating rotor, including the assumptions made for the W analysis.

RESPONSE

A reactor coolant pump lochd rotor as opposed to a f ree rotating rotor results in a larger flow resistance in the reactor coolant loops. Westinghouse assumes RCP coasting down and pump behavior calculated accordingly. During reflood portion of the transient locked rotor resistance is assumed for conservatism.

440.218 Please provide a nodalization diagram for large break and small break 15.6 analyses, if not already provided in response to a prior question, and include major characteristics such as flow length, flow area, and volume for each node.

1

RESPONSE

Figure 440.218-A attached shows the SATAN nodalization diagram for the large break analysis. The WREFLOOD noding scheme is identical to that used for standard plant analyses. Table 440.218-A includes volume for each node and flow area for each flow link.

WAPWR-PSSS 440-145 AMENDMENT 1 2854e:ld MAY,1986

{

The nodalization scheme used for the APWR Small Break LOCA Analysis with WFLASH is the same as the standard Westinghouse model preserted in WCAP-8970. A copy of the nodalization diagram is attached as Figure 440.218-B. Node dimensional data is shown in Table 440.218-B.

440.219 Please address the hot wall delay time in SP/90 as ccatrasted to an 15.6 older design W four loop plant. See also 440.197.

RESPONSE

The hot wall delay time for the SP/90 plant analysis is longer than that for older Westinghouse designs as expected. See W response to 440.197.

440.220 (15.6-12, second paragraph) "No fuel clad damage occurs." Please 15.6 discuss this in light of the temperature that is achieved and the generation of hydrogen that is stated to have occurred.

RESPONSE

Attachment 440.220 is provided to amend Page 15.6-12 of Module 1 to state that no fuel rod burst was predicted to occur. The peak clad temperature was predicted to occur early during blowdown when the clad experiences a high ramp heat up rate. During reflood at low heat up ramp rates (where standard plant rod burst is predicted to occur) the clad temperatures are low enough such that rod burst was not predicted to occur.

440.221 (15.6-12, bottom) The only parameters presented for the small break 15.6 results are listed as RCS pressure, core mixture height, and core power af ter reactor trip. Please provide vessel collapsed liquid level and the liquid level in other parts of the RCS if that level differs by more than one foot from the vessel level, and the net mass in the RCS as a function of time.

O 440-146 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY, 1986

RESPONSE

d For the 3", 4.313" and 6" diameter small break LOCAs presented in Section 15.6.4, the following additional parameters are provided in Figure 440.221-A through 440.221-U.

o RCS inventory o Vessel collapsed liquid level o Upper head mixture level o Loop seal mixture level O o Mixture level in the downhill (cold leg) side of the steam generator tubes o Pressurizer mixture level o Downcomer mixture level 440.222 Describe all deviations from prior EM analyses for a recent four loop 15.6 W plant, such as changes in nodalization, initial conditions, opera-tor actions, or modification to the computer codes. The plant selected for comparison should be one for which staff approval of the EM analyses is essentially complete. Modification of computer codes need not be described if such modifications are fully described in the references cited in Section 15 of Module 1. See also 440.197 and 440.223.

RESPONSE

As stated in response to question 440.216, the SP/90 analysis uses the assumptions in recent four loop Westinghouse plants as a basis.

Small deviations from prior EM analyses do exist due to differences O in APWR design. For example, upper plenum noding is slightly different due to the upper internals design (although the total number of vessel nodes is the same), two nodes are added for the CRT's, the injection point for the CRT's and SI pumps is in the O downcomer, and higher temperature ECCS water was assumed.

O v

440-147 AMENDMENT 1 WAPWR-PSSS 2854e:1d MAY, 1986

l l

]

440.223 A number of assumptions are implicit in the LOCA analysis and a number 15 6 of unsubstantiated assumptions are presented. Insofar as not previously addressed, please note the following items f rom SRP page 15.6.5-5 and provide an appropriate response:

"2. An adequate failure mode analysis has been performed to justify the selection of the most limiting single failure. This analysis is reviewed in part under SRP Section 6.3. If the design has been changed from that presented in previous applications, changes in the reactor coolant system, reactor core, and ECCS are reviewed with respect to the most limiting single failure.

"3. A variety of break locations and the complete spectrum of break sizes were analyzed. If part of the evaluation is done by referencing earlier work, design differences (ECCS, reactor coolant system, reactor core, etc.) between the facilities in question are reviewed. If there are significant differences, sensitivity studies on the important parameters should have been made by the applicant. If such sensitivity studies are not presented in the SAr't, the reviewer requests that they be made.

"4. The parameters and assumptions used for the calculations confonn to those of the approved evaluation model and were conservatively chosen, including the following points:

a. lhe initial power level is taken as the licensed oure thermal power for the number of loops initially assumed to be opera-ting plus an allowance of 2% to account for power measurement uncertainties, unless a lower power level can be justified by the applicant. The number of loops operating at the initiation of the event should correspond to the operating condition which maximizes the consequences of the event.

i O

WAPWR-PSSS 440-148 AMENDHENT 1 2854e:ld MAY, 1986

b. The neximum linear heat generation rate used should be based O' on 102% of the proposed licensed core thermal power and the technical specification limits on maximum linear heat generation rate.

O c. All permitted axial power shapes, as given in Section 4.3 of Normally, the the SAR, should be addressed by the analyses.

evaluation model will identify the least favorable axial shape as a function of break size. If the evaluation model

{'-- did not discuss axial shapes, or the discussion is not applicable to a given case, sensitivity studies are requested.

d. The initial stored energy was conservatively calculated by the applicant. The value used is checked against the applicant's steady-state temperatures, as given in SAR Section 4.4, similar calculations performed by the staff, or calculations done for similar plants by previous applicants.

I

e. Appropriate analyses are presented to support any credit taken for control rod insertion.

f. The applicant's analysis conservatively addresses the operation of the reactor coolant pump."

Note that many other portions of the SRP also are applicable.

RESPONSE

O 2. Because pumped safety injection in the ISS design is provided by four independent high head safety injection pump subsystems where:

{

- two of four HHS1 pumps are powered by each one of two O- independent electrical trains O

440-149 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

= _ _ _ _ , _ _ _ _ _ _ _ . . - - . _ -_

- no valve actuations are required in order for the ISS to perform its safety injection function and no injection to recirculation "switchover" is required.

- each HHSI pump delivers directly to the reactor vessel via four separate direct vessel injection lines:

the only active failures which could affect the safety injection performance of the ISS are the failure of a HHSI pump to start or run, or the failure of 2 of 4 HHS1 pumps due to an electrical power train failure or a start signal train failure, or support system train failure. (Since only active failures are assumed in the analyses failure of the accumulators and CRT's are not considered).

Because of this " simplicity of design", no failure mode and effects analysis of the ISS for its safety injection function was provided in Chapter 6.

3. Sensitivity studies have been performed for a variety of break locations, sizes, and types. Although these results are not presented in the SAR, the results can be presented to the staff possibly in the form of a report documenting the results of all sensitivity studies performed.

4. a. The initial power level assumed is confirmed to be 102 percent of licensed core thermal power with four loops in operation.

b. The maximum linear heat generation is based on 102 percent power and maximum core peaking factors.

O

c. Cases presented in this section assumed chopped cosine power shape. Sensitivity studies were performed for several O

440-150 AMENDMENT 1 WAPWR-PSSS 2854e:ld MAY,1986

i I l limiting power shapes. These power shapes included both top and bottom skewed shapes lying within the F g envelope. The results can be presented to the staff.

4

d. Initial stored energy was conservatively calculated using O methodolog! approved for standard plants. Fuel temperatures assumed are the highest calculated for the fuel life.

e. No credit is taken for control rod insertion for large breaks greater than 3.0 ft .

j

. f. It is confirmed that analyses conservatively address the operation of the RCP. See response to question 440.199.

440.224 (15.6-16, second paragraph) "Each of the four ISS subsystems is con-15.6 tained in a separate containment pressure pump (CPPE) compartment adjoining the reactor containment. These pump compartments are i designed to minimize the release to the environment of any radioac-

tive material resulting f rom leakage in the compartments, and return it to the containment. Therefore, radiological consequences from this source are virtually eliminated." This statement in regard to provision of the CPPE appears to contradict earlier statements which

! did not appear to be definitive, as well as recent information that the CPPE's are not in the design. Please provide the impact of CPPE l non-existence with respect to the conclusions formed in this section and correct the description of the plant.

RESPONSE

The CPPE's are designed to minimize the release of radioactivity to the environment resulting f rom the leakage of recirculating emergency core cooling solution outside containment.

!O

[

)(a,c) 440-151 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d i

440.225 (15.6-21. Table 15.6.4-1, Input Parameters Used in the ECCS Anal / sis) 15.6 Are the words " nominal" and " minimal" correctly applied in this table?

RESPONSE

Attachment 440.225 has been updated to reflect the analysis assumptions that the accumulator and core reflood tank values for pressures and water volumes should be minimal.

440.226 (15.6-22. Table 15.6.4-2, LARGE BREAK - TIME SE0VENCE OF EVENTS) 15.6

a. The end of ECC bypass and the end of blowdown times are given in the table as identical for the C = 1.0 column. Is this D

correct in light of the values being different in all of the other columns?

b. The accumulators are shown as emptying significantly af ter the bottom of core recovery. Please discuss the injection of nitrogen effect for the longer vessel and significantly different upper vessel internals as contrasted to typical W plants of existing design.

PESPONSE:

a. This is correct. It is not uncommon for end of blowdown conditions to be reached before end of bypass. The calculation cannot end for a DECLG before end of bypass is reached.

b. The effect of N gas injection from the accumulators has not 2

been looked at in great detail for the APWR. It is believed that the conclusion reached for the effect on standard plants is applicable for the APWR. The earlier study looked at the solubility of N in water, solubility of N in electrolyte 2 2 solutions, effect of N g gas on loop flow, effects of N 2 "

downtomer water level, and entrainment effects. The biggest ef f ect was found to be downtomer pressurization which tends to O

WAPWR-PSSS 440-152 AMENDMENT 1 2854e:ld MAY, 1986

\

l increase core flooding rates. Since its effect tended to be an

< overall benefit neglecting the effect of N 2 gas appears to be l conservative. For more detail please see WCAP-8471.

440.227 (15.6-23 Table 15.6.4-3, LARGE BREAK RESULTS) What material is 4 O 15.6 included in the calculation of the metal water reaction percentages?

Please address your response to both the actual materials in the f

reactor vessel and the basis for the percent as contrasted to the 10CFR 50.46(b) rule. (See 440.195.)

RESPONSE

Material included in the calculation is the total zircaloy in the

]

j cladding of all the fuel rods in the active region of the core.

440.228 (Figure 15.6.4-4, Peak Clad Temperature, Elevation 6.5 Feet) The tem-15.6 perature from roughly 160 seconds appears to be in excess of 300'F.

Is this correct? If so, why, since the water is significantly cooler?

RESPONSE

l This is correct since lower plenum water (core inlet) temperature is l

in excess of 250*F from lower internals metal release and core decay

heat is still being generated by the fuel rods. The metal release is based on a lumped heat transfer model (further described in t

WCAP-8471) which releases heat exponentially based on decay constants j for the metal. This is conservative to a slab heat transfer model.

l The core pressure is on the order of 50 psi at this time.

O 440.229 [ Figure 15.6.4-16, Accumulator Flow (Blowdown)) What is the reason 15.6 for the decrease in accumulator flow rate that is shown in the vicinity of 24 seconds?

RESPONSE

Accumulator flow decreases by approximately ]1b/sec.,whichisa ta,c) small change, and is caused by a small pressure increase in the cold leg - probably due to the injection of the core reflood tanks.

O WAPWR-PSSS 440-153 AMENDMENT 1 2854e:ld MAY, 1986

440.230 [ Figure 15.6.4-32, Core Mixture Height (3 inch)] What is the reason 15.6 for the perturbation near 900 set (which also occurs in the RCS pressure curve) and for the increase that occurs at 1500 sec? In regard to the 900 second pertuibation the staff notes this also occurs for the four inch and six inch breaks, with a change that is significantly greater for the six inch break.

RESPONSE

The perturbations in core mixture level for the 3 inch break at 900 seconds, the 4 inch break at 370 seconds and the 6 inch break at 170 seconds are caused by the loop seal uncovering. Imediately af ter the loop seal is uncovered, steam generated in the core has an unimpeded vent path through the RCS and out of the break. Pressure in the core rapidly decreases and a surge of water into the core from the vessel downcomer pushes the core mixture level up. (See additional figures of loop seal and downtomer mixture levels provided in Question 440.221.)

During the 3 inch break, the perturbation in core mixture level at 1500 seconds is caused by the small amount of liquid remaining in the l

upper head draining into the core and upper plenum node. See the figure of upper head level provided in Question 440.221.

440.231 (Figure 15.6.4-33, Core Power Af ter Reactor Trip) What is the meaning 15.6 of " Total residual heat (with 4 shutdown)"?

RESPONSE

Attachment 440.231 is provided to amend Figure 15.6.4-33 to correctly read "(with 4% shutdown) consistent with standard Westinghouse SAR format.

440.232 What is the status of the pressurizer heaters for purposes of the Ch.

15.6 15 analyses? For each of the events, show that a conservative assumption has been used.

O WAPWR-PSSS 440-154 AMENDMENT 1 2854e:ld MAY, 1986

l

! RESPONSE:

Operation of pressurizer heaters is not assumed in any of the Chapter 15 analyses. This is because the effect of their operation is either  !

negligible or nonconservative.

.O For both large and small break LOCA transients presented in this

module, their contribution is negligible due to the rapid rate of decrease of pressure and the rapid emptying of the pressurizer.

O 440.233 (1.2-5, '.hird paragraph) "The CVCS letdown heat exchangers, located 1.2 inside the containment would permit the letdown flow to be subcooled

! before it is released into the EWST."

a. Will the P&I0 provided with the CVCS module show this flow path,

> including all valves and interconnections?

b. Will the CVCS module contain suf ficient information to establish i that the letdown flow will be subcooled as stated here?

i RESPONSE:

! Module 13 " Auxiliary Systems" will contain detailed P& ids of the CVCS which will show its interconnections with other systems like the

! RCS and the ISS. The P&ID will show all the valves including their type and the instrumentation. It should be noted that as mentioned i

l

! in our response to question 440.193 that the emergency letdown line will not share any of the CVCS equipment but instead will be two l separate lines each of which is connected to a RCS hot leg and is routed to the EWST. Each emergency letdown line contains two f ail closed isolation valves and an orifice. The emergency letdown is sparged into the EWST to minimize the subsequent containment cleanup.

440.234 (1.2-5 last paragraph) Will the module that covers the diesel ener-

! 1.2 gized emergency electrical power trains identify all items connected to each train and provide a comparison to train capacity?

O 440-155 AMEN 0 MENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

RESPONSE

Table 8.3-2 of Module 9 "I&C Electrical Power", provides all the information requested here.

440.235 In regard to emergency power trains, please contrast time between receipt of a start-up signal and time of loading the diesels to the requirements for this timing. Include:

a. A comparison to regulatory requirements using SP/90 safety related equipment. For example, the SP/90 is equipped with core reflood tanks which provide flow for an extended time in a large break LOCA condition. This may make operation of pumps unnecessary for several minutes af ter accident initiation. If this is correct, and rapid start requirements for diesels are based on pump need within a few seconds of initiation of the accident, then one might relax start-up time requirements to obtain more reliable, cost effective, emergency power trains.

b. A comparison to the most limiting design basis accident (s) with timing based upon best estimate plant response.

c. An assessment of the relationship between startup time, cost, and reliability of the diesel energized emergency electrical power systems.

RESPONSE

a. Suf ficier.t analysis has been performed to date to justify a l

relantion of the 10 second diesel generator start-up time requirement. However, the sensitivity studies necessary to establish a definite defendable extended startup time have not been performed. These analyses would be required for all postulated accidents that rely on emergency electrical ons site power from the Diesel Generators.

O 440-156 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:ld

b. Analysis has not been performed to date that would provide a comparison for the Diesel Generator start-up times for the most limiting design basis accident and the best estimate plant response to that accident.

c. Westinghouse has not performed an evaluation of the cost reduction and reliability improvements associated with a relaxed 0/G start-up time. However, it is expected that a significant relaxation in start-up time would be required in order to obtain any benefits in cost or reliability.

440.236 (1.2-6, third paragraph) "In the event of a steam break accident the 1.2 high head pumps inject borated water ... with suf ficient shutdown reactivity to counteract any reactivity increase caused by the resulting system cooldown and to compensate for the change in RCS volume. The high head pumps would continue injecting ... to prevent I the possibility of a return to criticality." Please discuss the l implications of a steam break inside containment on boron concen-tration in the EWST and the result of the dilution upon criticality I considerations.

Provide specific numerical values of the pertinent parameters such as the volume of water added to the EWST and the boron concentration that results. Also include consideration of mixing (or lack of mixing) in the EWST, and the location of suction inlets.

I

RESPONSE

)

The postulated occurrence of a steam line break (or feedline break) inside containment, will not af fect the boron concentration of the EWST such that the capability of the ISS to borate the RCS and provide required shutdown is prevented.

The SP/90 containment is arranged so that unborated water that is

spilled / released into the containment following a postulated steam line break will not return directly to the EWST. Referring to 440-157 AMEN 0 MENT 1 WAPWR-PSSS MAY,1986 2854e:1d

e Module 3 " Introduction and Site", Figure 1.2-2, the spillways back into the EWST (12, 24-inch diameter pipes) are elevated above the floor in the " annulus" between the loop compartments and the containment wall. The floor of the annulus is located at a higher elevation than the loop compartment floor. Also, curbs are provided in the loop compartment to direct any water on the loop compartment floor to the room below the upper internals storage pit where sump pumps are provided. This room, which is lower _than the loop compartment floor elevation, has a volume of _

The amount of water which would be released or spilled following a steam line break (SLB) has been conservatively determined to be ta,cl of water. This includes: 1) the maximum inventory of one steam generator (SG) at hot zero power (limiting case); 2) the mass of steam from the other 3 SG's and steam lines released before steam line isolation is completed; 3) main feedwater flow to the faulted SG until isolation is completed; and 4) emergency f eedwater flow to the f aulted SG until isolation by the operator at 30 minutes. Thus it is clear that the total mass of unborated water / steam that could be released or spilled f rom the break would be directed to and could be accommodated in the compartment below the upper internals storage pit.

If it is assumed that containment spray is actuated following the

- a (a,c) postulated St.B. borated water will be introduced into the containment from the EWST via the RHR/CS pumps-spray headers at I"'*3 gpm (maximum safeguards) or _ _

gpm (minimum safeguards). The borated spray would mix with the water / steam released f rom the break, sequentially filling the compartment below loop the upper internals pit, the upper internals pit, the compartment floor area, the reactor vessel cavity, incore instrument tunnel and other lower compartments, and finally the annulus area outside the loop compartments. The water level would continue to rise until the water level reached the elevation of the EWST O

440-158 AMENOMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

]

, spillways. The amount of water required to flood the containment to ,

this elevation is calculated to be gallons (i.e. _ ] (a,c) gallons of borated spray and _ _

gallons of unborated water f rom the SLB), and would begin to spill back into the containment at ~30 minutes to 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> af ter spray initiation.

It is clear from the above discussions that:

- the water released f rom a postulated SLB, even with centainment j

spray actuation, will have no impact on the boron concentration of EWST water injected into the S by the HHSI pumps to counteract the reactivity increase caused by system cooldown and

' coolant shrinkage which occurs immediately following the break.

l

- for postulated SLB's which do not actuate spray or if spray is assumed to be terminated at or before 30 minutes, there is no impact on EWST boron concentration.

ta,c) l

\

- the boron concentration of water returning to the EWST can be conservatively assumed to be completely mixed .

ta.cl j ,

In order to complete accident recovery following a SLB, the operator O would proceed to borate the RCS to cold shutdown concentration prior to initiating RCS cooldown using the intact steam generators. This boration of the RCS is accomplished by injecting borated water f rom the EWSi using the HHSI pumps while " letting down" RCS coolant to the O EWST via the emergency letdown line.

O 440-159 AMEN 0 MENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

The water returning to the EWST from the containment floor enters the EWST via 12, 24-inch diameter pipes which are located in four separated locations and are 130' f rom the nearest HHSI and RHR/CS pump suction lines. The continued operation of the spray pumps will serve to mix the 2500 ppm EWST water with the 2045 ppm water on the containment floor. Assuming the most limiting case - that the EWST water has been well mixed with the water above the containment floor (i.e.12250 ppm) to minimize its boron concentration - the RCS can be borated to its cold shutdown concentration (0 - 1300 ppm used as limiting boration requirement) in 7.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. This capability is consistent with WAPWR plant design bases, and with the stated capability to reach cold shutdown to 200'F within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

The postulated SLB case described above provides the worst possible scenario for the di.lution of the EWST by secondary side water (i.e.

non LOCA case) where boration of the RCS is required prior to achieving safety grade cold shutdown. This considers the postulated inadvertent opening of the steam generator overfill protection system (until the line is closed by the operator), the postulated failure of the SG 1evel control system which would result in reactor trip and main water system isolation (protection grade instrumentation) prior to initiation of the SG overfill protection system function, or SG j

tube rupture where the SG inventory is mixed with RCS fluid and the SG inventory is retained in the steam generator.

140.237 (1.2-10, second paragraph) "The emergency boration/ letdown is accom-1.2 plished by initially depressurizing the RCS by temporarily opening i

pressurizer vent valves untti the RCS pressure is below the shutoff head of the high head pumps." Please provide the timing and flow rates for this process.

RESPONSE

The operator would initiate emergency boration of the RCS by starting two of the four HHS1 pumps. Since the shutoff head of these pumps is

-1785 psig, no injection of borated water will occur and the HHS1 O

WAPWR-PSSS 440-160 AMEN 0 MENT 1 2854e:1d MAY, 1986 l

pumps will simply run at their miniflow flowrate. When satisfactory operation of the HHSI pumps is confirmed, the operator will initiate depressurization of the RCS by opening one pressurizer PORV and reduce the RCS pressure from 2250 psig to 1600 psig. Assuming normal pressurizer level at the initiation of the depressurization, this depressurization would take ~120 seconds, at which time the two operating HHSI pumps would be injecting 220 gpm into the RCS. This HHSI flow will match the letdown flow f rom the RCS hot leg to the EWST through one of the two emergency letdown flowpaths.

The operator actions and times for initiating emergency boration operation will change somewhat depending on the assumed pressurizer pressure and level initial conditions. For example, if the pressurizer level is low (post reactor trip level), the

! depressurization to 1600 psig will take longer than two minutes and the operator would delay opening the emergency letdown flowpath until the HHSI pump injection has increased pressurizer level.

10

440.238 (1.2-11, last paragraph) "The operator would first open the safety l

1.2 grade pressurizer vent system, th,en start one or more of the ISS high head pumps. Af ter the feed and bleed operation was established and the pressurizer water level confirmed. . . . ." Please explain how the operator determines pressurizer water level with respect to a meaningful parameter with the vent valves open.

RESPONSE

Following the initiation of the emergency feed and bleed operation, O core cooling is provided by boiling in the core region with steam being vented via the pressurizer PORV's to the PRT, and sparged into the EWST. During this method of heat removal, meaningful pressurizer level indication will not be available and the operator would rely on O core exit thermocouples and hot leg temperature indication, RCS pressure indication, reactor vessel level indication, and HHS! pump flow indication to assure that adequate core cooling was being maintained.

440-161 AMENDMENT 1 WAPWR-PSSS 2854e:1d MAY, 1986

i 1

440.239 (1.2-11, last paragraph) "The flow rate through the CVCS letdown heat 1.2 exchangers would be controlled by the operator to ensure that subcooled flow returned to the EWST." Please provide a description of the instrumentation and controls that will be available to the '

operator for this process. See also 440.233.

RESPONSE

As discussed in the response to question 440.193, the CVCS will not provide emergency letdown; insteed, the ISS will h:ve two separate lines from the RCS hot legs to the EWST. These ISS lines will be shown in the ISS information (description and flow diagrams) for the integrated submittal in 1986. A sketch of the ISS (Figure 440.239-A) and CVCS (Figure 440.239-B) are attached for your information. Each ISS emergency letdown line has two series solenoid valves that are normally closed fail closed. These lines also have one orifice to limit the tlow to 250 gpm at normal RCS pressures. The operator has control of these valves in the control room as well as valve position indication.

440.240 (Table 1.0-2) Please consider the following if not addressed in 1.8 response to prior questions:

a. "The vapor pressure of the water in the Emergency Water Storage Tank (EWST) is assumed to be equal to the containment pressure for the long term post-accident recirculation. The vapor pressure of the EWST water cannot exceed the containment total pressure; therefore, assuming they are equal gives the limiting low value of available NPSH." Consider a situation in which containment pressure has been elevated, and then reduced. One can encounter c situation ir, which the vapor pressure is in equilibrium throughout the tank, with temperature gradients present. Now the vapor pressure at lower elevations is greater than that of the containment. While the staf f recognizes that this condition is not long term unless there is a concentration gradient which contributes an ef fect, it would appear that this l

440-162 AMENDMENT 1 l WAPWR-PSS)

MAY, 1986 l 2854e:1d l

l t

l' i

could occur in short term operation. Please comment, and address the W statement in regard to the assumption being a limiting assumption. Include consideration of the suction piping and pump j inlet regions during containment depressurization.

b. Please address the presence of significant dissolved gases and i

other materials, such as oil, on the behavior,

c. Please discuss the influence of non-uniformities in the EWST water, such as may exist due to injection of low boron content water from the steam generators or the pressurizer relief tank overflow.

RESPONSE

! Table 1.8-2 page 1.83 contains the Westinghouse position on R.G.1.1, which was primarily written to preclude post accident containment

]'

pressure from being used as a means for satisfying NPSH requirements j for ECCS and containment spray pumps during the post accident recirculation phase. For compliance with R.G. 1.1, Westinghouse

~]

bases all NPSH calculations for all Westinghouse PWRs on the conservative assumption that the water in the containment sump would reach a maximum uniform temperature equal to the saturated temperature corresponding to the containment pressure.

The staf f's postulated scenario, where the containment pressure has been elevated and then reduced, would obviously produce boiling in the containment sump water, if the water were indeed at a saturation O temperature corresponding to the maximum containment pressure. This postulated scenario is applicable, not just to WAPWR, but to all PWRs and BWRs that are forced to make the worst case assumption that the containment water is saturated. The actual conservatively calcu-O lated surry water temperature is significantly less than saturation.

In order to cla r,1 f y the thermal behavior of the water in the .

containment and the! EWST for the WAPWR, subsequent to a large LOCA, O the following discussion is provided.

WAFWR-PSSS 440-163 AMEN 0 MENT 1 i 2854e:1d MAY, 1986

The EWST normally contains 580,000 gallons of 100'F water. In the event of a large LOCA and no single failure, the four high head pumps would take suction f rom the EWST and deliver a total of 4000 gpm to the RCS downcomer. If the containment spray system has been initiated by a high containment pressure signal, the four low head pumps would deliver a total of 14,000 gpm to the containment spray ring headers.

The total flow from the EWST, based on the maximum runout flowrates of the four high head pumps (1000 gpm each) and the four low head pumps (3500 gpm each) would be 18,000 gpm. The 14,000 gpm containment spray water would immediately start to collect in the lowest levels of the containment building, such as the reactor vessel cavity, and cool the saturated RCS water that spilled to the containment floor during the initial blowdown phase.

In the event of a cold leg LOCA, a large percentage of the high head pump flow would start to spill to the containment floor as soon as the downcomer was filled. Af ter the core was recovered, the majority of the high head pump flow would spill to the containment without any significant increase in temperature above the original 100*F EWST water temperature.

With regard to the containment water temperature, the worst break location would be a hot leg break because all of the 4000 gpm water injected into the Direct Vessel Injection (DVI) nozzle would pass through the core before spilling to the containment. Therefore, the saturated water spilling f rom the hot leg break would mix with the -

14,000 gpm subcooled water f rom the containment spray headers.

3 The total dead volume below the EWST spillway is 1507.2 m . If it is assumed that this dead volume is primarily filled by the 18,000 gpm being drawn f rom the EWST, it would take approximately 21 minutes before any water would return to the EWST via the EWST spillways.

O 440-164 AMENDHENT 1 WAPWR-PSSS 2854e:ld MAY,1986

i*

l j

If a loss of offsite power and a failure of one of the two emergency l

l diesel generators is assumed to occur simultaneous with the postulated LOCA, only twc' high head pumps and two low head pumps would operate. For this case, the maximum draw down rate for the EWST would only be 9,000 gpm, therefore, it would take 42 minutes l before any water would return to the EWST.

i When the water on the containment floor reaches the top of the EWST f

spillways and begins to spill back into the EWST, approximately

[ The j 250,000 gallons of 100*F water would still remain in the EWST.

location of the six spillways and the four sump pits in the floor of the EWST ensure total mixing of the water returning to the EWST with

[ the water remaining in the EWST.

440.241 In anticipation of work associated with the SP/90 PRA and associated system response:

6

a. Some discussion has been presented pertinent to vertical and O horizontal pumps for PSSS components. Please provide the considerations which led to the SP/90 design, and include a j discussion of expected response under severe accident conditions 2

where the pumps may be used under conditions which extend beyond l

the design basis, s

! b. One of the situations that will be encountered in the PRA, and an

! item that will be considered in the review outside of the PRA, is

! ATWS. With respect to this topic, please discuss emergency boration under ATWS conditions.

RESPONSE

a. The selection of a spherical containment configuration for the O WAPWR essentially dictated the selection of horizontal pumps for PSSS components. The compartments beneath the concrete support cradle, at elevation 72.0 meters, are ideally suited for the horizontal pump application. The higher NPSH requirement for the O

440-165 AMENDMENT 1 WAPWR-PSSS MAY, 1986 2854e:1d

horizontal pumps is satisfied by the elevation difference between

• j the EWST, at elevation 79.4 meters, and the centerline elevation '

of the pumps.

In the event that a cylindrical containment were to be utilized for the WAPWR, the vertical pump configuration would be reconsidered if it provided a benefit to the construction of the adjoining safe. guards building.

b. Module 16, which has recently been submitted to the NRC shows that for ATWS the reactor must be shut down within 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> by manual rod insertion or by boration by the CVCS. The CVCS is used because the RCS will tend to be at elevated pressures and the higher head of the charging pumps may be necessary to inject boric acid. Also, because ATWS is not a design basis accident, the use of a non-active, tech spec. system is acceptable.

O O

l l

l O

l WAPWR-l'SSS 440-166 AMENDMENT 1 2054e:ld MAY, 1986

l 450.1 (6.5.2) What is the basis for assuming 90,000 cfm mixing between the j 80% of the containment volume that is sprayed and the 20% that is

! not? Identify the locations of the unsprayed volume relative to a

i likely leakage paths.

l 1'O V RESPONSE:

90.000 cfm is the minimum post accident containment fan cooler flow i

rate. This flow rate is conservatively assumed to be the only

, communication between the sprayed and unsprayed containment regions.

l Both the sprayed and unsprayed regions are assumed to leak at the l l containment design leak rate.

! 450.2 (6.5.2) What is the pH of the EWST as a function of time during the

! first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the LOCA analyzed in 15.6.57 How does the Westinghouse model consider the basicity of alkali metal and alkaline earth fission products that might be washed into the EWST.

j I

l RESPONSE: '

Initially the pH of the EWST will be approximately 4.5 to 5 i'

~

Well within the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, caustic will be i .

i added to the sump solution to raise the pH into the range of 7 to 9.5. The dissolution of alkali metals (Cs and Rb) is not considered j in the pH calculation. For the design basis LOCA (TID 14844 source)

I these metals are assumed to exist in small quantities (1% of core inventory is released) and would have little impact on the pH of the EWST solution, i

450.3 (6.5.3) Describe the safeguard area cleanup system and list the l equipment contained in the volumes it serves.

I 1

I RESPONSE:

The safeguard area cleanup system will be described in Module 10,  ;

j " Containment Systems".

lO 450-1 AMENDMENT 1 WAPWR-PSSS 2854e:1d MAY, 1986

(

- _ . - , - - - . _ _ . - - , - , ~ - - - . ~ . - - . . - - . - - - - - . _ . , . - . _ - . . - _ - - .

450.4 (15.6.2) Are all primary sample and instrument lines outside j containment completely within volumes served by the safeguard area cleanup system?

RESPONSE

l No.

l l

450.5 (15.6.5) Are all lines drawing from the FWST that penetrate containment completely within volumes served by the safeguard area cleanup system?

RESPONSE

Yes.

O l

l l

O O

O WAPWR-PSSS 450-2 AMENDHENT 1 2054e:ld MAY, 1986

1 ATTACHMENT 440.40 5.0 REACTOR COOLANT SYSTEM AND CONNECTED SYSTEMS 5.4 COMPONENT AND SUBSYSTEM DESIGN 5.4.7 Residual Heat Removal System The functions performed by a conventional nuclear plant residual heat removal (RHR) system are integrated within the integrated safeguards system (ISS) of the WAPWR. Those components within the ISS that perform an RHR function are the four low head pumps, four RHR heat exchangers, and the associated valves, piping, and instrumentation.

Following initial cooldown of the plant by steam dump to the main condenser, the four low head pumps would be aligned to recirculate reactor coolant through the core by taking suction from the RCS hot legs and returning the coolant to the reactor vessel through the four RHR heat exchangers. The four RHR pumps and four heat exchangers would be capable of reducing the RCS coolant temperature from 350*F to 150*F in less than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> following reactor shutdown. It should be emphasized that the ISS provides four RHR subsystems which would permit three of four subsystems to be taken out of service during long-term shutdown operation.

The low head pumps have multiple uses. During plant cooldown and refueling operations, they act as conventional RHR pumps. In their accident mitigation role, they act as containment spray pumps (see Section 6.2.2) or could be aligned to provide a long term ECCS function (see Section 6.3.2.2.2).

Additionally, the pumps are used to transfer refueling water from the s

O' refueling canal and the Emergency Water Storage Tank (EWST) at the beginning of refueling operations. Refueling water is returned to the EWST f rom the refueling cavity at the end of refueling operations by a gravity drain.

O The RHR heat exchangers perform the heat removal function during plant cooldown and refueling operations as well as during accident recovery operations (see Section 6.3.2.2.6). ]1 WAPWR-PSSS 5.4-1 AMENDMENT 1 4854e:1d MAY, 1986 l

i i ATTACHMENT 440.111 Water can be removed from the core reflood tanks by opening the appropriate valves in the test line and permitting flow to return to the EWST. Periodic checks of the core reflood tank boron concentration are made through the sampling system.

Redundant level and pressure indicators are provided on each core reflood tank

, with readouts on the main control board. Each indicator is equipped with high

! and low-level alarms. The margin between the minimum operating pressure or level and the maximum operating pressure or level provides a range of j acceptable operating conditions. The band width is sufficient to minimize the r frequency of adjustments in the core reflood tank (CRT) pressure or level requi[e8tocompensateforleakage.

The design parameters for the core reflood tanks are provided in Table 6.3-2.

l 6.3.2.2.5 Emeroency Water Storaae Tank (EWST) j The EWST is located at the lower elevation inside the containment building and provides a continuous suction source for the high and low head pumps, thereby eliminating the switchover from injection to recirculation. The EWST and loop compartments would be arranged to minimize the containment cleanup in case of 4

minor accidents, such as reactor coolant pump seal failure or instrument line breaks. Any discharge from the pressurizer relief tank should also be routed to this tank. The required water volume depends on the Refueling Canal Volume

]

.I which is not expected to exceed [ ] gallons. ta,el I

1 Analyses are perfonned to determine the minimum water level in the EWST during recirculation. These analyses consider the amount of water trapped in lower l

! containment compartments and the delay time for water to return to the EWST. l The design parameters for the EWST are provided in Table 6.3-2.

6.3.2.2.6 Residual Heat Removal (RHR) Heat Exchanaers 4

! Four RHR heat exchangers are provided, with one RHR heat exchanger assigned to

! each of the four subsystems. Each exchanger is sized to remove [ ] percent I*'CI l WAPWR-PSSS 6.3-14 AMENDMENT 1 4854e:1d MAY, 1986 l

ATTACHMENT 440.115

a. Where possible, packless valves are used.

b. Other valves which are normally open, except check valves and those which perform a control function, are provided with backseats to limit stem leakage.

c. Normally closed globe valves are installed with recirculation fluid pressure under the seat to prevent stem leakage of recirculated (radioactive) water.

d. Relief valves are enclosed, i.e., they are provided with a closed bonnet. j Motor-Operated Gate Valves The seating design of all motor-operated gate valves is of the crane flexible wedge design. These designs release the mechanical holding force during the first increment of travel so that the motor operator works only against the frictional component of the hydraulic unbalance on the disc and the packing box O friction. The dbcs are guided throughout the full disc travel to prevent chattering and to provide ease of gate movement. The seating surfaces are hard faced to prevent galling and to reduce wear.

Where a gasket is employed for the body to bonnet joint, it is either a fully trapped, controlled compression, spiral wound asbestos gasket with provisions for seal welding, or it is of the pressure seal design with provisions for seal welding. The valve stuffing boxes are designed with a lantern ring leakoff connection with a minimum of a full set of packing below the lantern ring and a O minimum of one-half of a set of packing above the lantern ring. A full set of packing is defined as a depth of packing equal to 1-1/2 times the stem diameter.

The motor operator incorporates a " hammer blow" feature that allows the motor to O impact the discs away f rom the backseat upon opening or closing. This " hammer blow" feature not only impacts the disc but allows the motor to attain its operational speed prior to impact. Valves which must function against system '

pressure are designed such that they function with a pressure drop equal to full 1 O system pressure the valve disc.

6.3-16 AMENDMENT 1 WAPWR-PSSS 4854e:1d MAY, 1986

1 ATTACHMENT 440.122 recirculation or operator error and maintain the performance objectives desired in subsection 6.3.1. Separate trains of pumps, heat exchangers and flow paths are provided for redundancy. The initiating signals for the ECCS are derived from independent sources as measured from process (e.g, pressurizer low pressure) or environmental (e.g, containment high pressure) variables. Redundant as well as functionally independent variables are measured to initiate the Safeguards signal. Each train is physically i

separated and protected where necessary so that a single event cannot initiate l a conunon f ailure. Power sources are divided into independent trains supplied from the separate emergency buses supplied f rom of f site power. Sufficient

{

} diesel generating capacity is maintained on site to provide required power to each train. The diesel generators and their auxiliary systems are completely l independent and dedicated to one or two of the trains.

The preoperational testing program ensures that systems, as designed and constructed, will meet the functional requirements. The ECCS is designed with the ability for on-line testing of most components so the availability and operational status can be readily determined. In addition to the above, the l

i integrity of the ECCS is ensured through examination of critical components I during the routine inservice inspection.

The reliability program extends to the procurement of ECCS components such a that only designs which have been proven by past use in similar applications l are acceptable for use. The quality assurance program as described in Chapter l 17 ensures receipt of components only after manufacture and test to the

applicable codes and standards.

i 1

1 6.3.2.5.1 Active Failure Criteria The failure of a powered component, such as a piece of mechanical equipment, component of the electrical supply system or instrumentation and control O equipment, to act on command to perform its design function is considered an active f ailure. Examples include the failure of a motor-operated valve to

) move to its correct position, the failure of an electrical breaker or relay to

respond, the failure of a pump, fan, or diesel generator to start, etc.

WAPWR-PSSS 6.3-20 AMEN 0 MENT 1 4854e:1d MAY, 198f3 i f

ATTACHMENT 440.145

a. The calculated peak fuel element clad temperature is less than 2,200*F

b. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircaloy in the reactor.

c. The clad temperature transient is terminated at a time when the core O geometry is still amenable to cooling. The cladding oxidation limits of 17 percent are noi exceeded during or af ter quenching.

d. The core remains amenable to cooling during and af ter the break.

The core temperature is maintained at an acceptably low value and O e. 1 decay heat is removed for an extended period of time, as required by the longlived radioactivity remaining in the core.

6.3.3.4 Major Secondary System Pipe Failure Discussion The steam release from a rupture of a main steam pipe would result in an increase in energy removal from the RCS causing a reduction of coolant temperature and pressure. In the presence of a negative moderator temperature coefficient, the cooldown results in an in::ertion of positive reactivity.

l There is an increased possibility that the core will become critical and return to power. A return to power following a steam pipe rupture is a potential problem; however, analysis demonstrates that the core is ultimately shut down by the injected boric acid.

Minimum capability for injection of boric acid (2,500 gpm) solution is assumed corresponding to the most restrictive single failure in the ECCS system. For the cases where offsite power is assumed to be available, the high head pumps are assumed to start immediately upon receipt of the "S" signal and to achieve full speed in five seconds. The water initially within the high head pump piping is assumed to be swept into the RCS (with no credit taken for its boron) before the 2,500 ppm water f rom the EWST reaches the core. For the O cases where offsite power is assumed not to be available, an additional 10 Q

second delay is assumed to start the diesels. The necessary ECCS equipment are then loaded onto the diesels according to the sequencer.

WAPWR-PSSS . 6.3-31 AMENDMENT 1 4854e:ld MAY, 1986

ATTACHMENT 440.148 See the " Secondary Side Safeguards" module for the results and conclusions of this analysis.

6.3.3.5 Steam Generator Tube Failure The accident examined is the complete severance of a single steam generator tube at power.

Assuming normal operation of the various plant control systems, the following r sequence of events is initiated by a tube rupture:

a. Pressurizer low pressure and low level alarms are actuated and charging pump flow increases in an attempt to maintain pressurizer level. On the secondary side, there is a steam flow /feedwater flow mismatch before the trip as the feedwater flow to the affected steam generator is reduce ( due to the additional break flow which is now being supplied to that unit.

t O b. The steam generator blowdown liquid monitor and the condenser offgas radiaton monitor will alarm, indicating a sharp increase in radioactivity in the secondary system and will automatically terminate steam generator blowdown.

t

c. Continued loss of reactor coolant inventory leads to a reactor trip signal generated by low pressurizer pressure or low DNBR reactor 1 trip. The "S" signal automatically terminates normal feedwater supply and initiates emergency f eedwater addition. After reactor trip, the O break flow reaches equilibrium at the point where incoming safety injection flow is balanced by outgoing break flow. The resultant break flow persists from plant trip until operator action is taken to bring the primary system and the f aulted steam generator secondary O system pressures into equilibrium,

d. The reactor trip automatically trips the turbine and, if offsite power is available, the steam dump valves open permitting steam dump to the

, O condenser. In the event of a coincident station blackout, the steam WAPWR-PSSS 6.3-32 AMENDMENT 1 4854e:ld MAY , 1986

1 ATTACHMENT 440.150(a)

An "S" signal normally results in a reactor trip followed by a turbine trip.

However, it cannot be assumed that any single fault that actuates ECCS will also provide a reactor trip. If a reactor trip is generated by as spurious signal, the operator should determine if the signal was transient or steady state in nature and if the safety injection signal must be blocked. For a spurious occurrence, the operator would terminate the safety injection and If the SIS actuation 1 maintain the plant in the hot shutdown condition.

instrumentation future plant operation would be in accordance with the Technical Specifications.

Conclusions Results of the analysis show that spurious ECCS operation without immediate reactor trip presents no hazard to the integrity of the RCS.

6.3.4 Tests and Inspections 6.3.4.1 ECCS Performance Tests Preliminary operational testing of the ECCS can be conducted following flushing and hydrostatic testing, with the system cold and the reactor vessel i head removed. Subsequent system performance testing can be conducted during each major fuel reloading operation with each subsystem aligned to take suction f rom the EWST and to deliver to the EWST via the system test line.

Each pump can also inject into the reactor vessel, with the overflow from the .

1 reactor vessel spilling into the refueling canal. Simultaneously, the safety '

injection block switch is reset and the breakers on the lines supplying off site power are tripped manually so that operation of the emergency diesels is tested in conjection with the ECCS. This test should provide information including the following facets:

O

a. Satisf actory safety injection "S" signal generation and ' transmission.

b. Proper operation of the emergency diesel generators, including sequential load pickup.

WAPWR-PSSS 6.3-35 AMENDMENT 1 4854e:ld MAY , 1986

ATTACHMENT 440.150(c) l

c. Pump starting times.

I

d. Pump delivery rates. l Separate flow tests of the low head and high head pumps should be conducted l during any system performance test operation to verify the pump head / flow characteristics. In addition, these tests are required to establish / verify flows in conjunction with the required pump discharge flow rates for both the reactor vessel injection and hot leg injection modes of operation. During O these tests, the pumps are aligned to take suction f rom the EWST and to  !

discharge into the reactor vessel through the injection lines. More l

specifically, the system performance tests are required to ensure that the appropriate. sized flow-restricting orifice plates are installed in the high head pump miniflow lines and the high head and low head pump discharge headers.

Each accumulator and core reflood tank is filled with water from the EWST and l1 pressurized with nitrogen with the motor-operated valve on the discharge line closed. The valve is opened and the accumulator and core reflood tank allowed-O to discharge into the reactor vessel as part of the operational startup testing with the reactor vessel head off.

6.3.4.2 Reliability Tests and Inspections Routine periodic testing of ECCS components and all necessary support systems at power is planned. Valves which operate af ter a LOCA are operated through a complete cycle, and pumps are operated individually in this test on their ,

miniflow lines.

If such testing indicates a need for corrective maintenance, the redundancy of equipment in these systems permits such maintenance to be performed without shutting down or reducing load under certain conditions. These conditions include considerations such as a period within which the component should be restored to service and the capability of the remaining equipment to provide the minimum required level of performance during such a period.

WAPWR-PSSS 6.3-36 AMENDMENT 1 4854e:1d MAY, 1986

ATTACHMENT 440.151 The signals that are generated by the protection logic and used to initiate the "S" signal are the following;

a. Pressurizer trip signal, produced by two-out-of-four (2/4) pressurizer low-pressure signals

b. Hi-1 containment pressure trip signal, produced by two-out-of-four (2/4) containment Hi-1 pressure signals O c. Steam line low-pressure signal, produced by two-out-of-four (2/4) steam line low-pressure signals in one line

d. Excessive cooldown, produced by low T-cold signals in two-out-of-four (2/4) loops coincident with a neutron flux of below 10 percent

e. Manual safety injection actuation from the control board The actuation signal that initiates containment isolation phase A and containment ventilation isolation is referred to as the "T" signal. The "T" signal is initiated from the same protection logic signals that produced the "S" signal, except that a separate nanual actuation switch is provided on the control board that permits the operator to initiate containment isolation phase A actuation without initiating the ECCS. In addition, the "S" signal reset is separate from the "T" signal reset.

The actuation signal that initiates spray actuation and containment isolation phase B is referred to as the "P" signal. The signals that are generated by the protection logic and used to initiate the "P" signal are the following:

l

a. Hi-3 containment pressure trip signal, produced by two-out-of-four (2/4) containment Hi-3 pressure signals O

b. Manual actuation from control board WAPWR-PSSS 6.3-38 AMENDMENT 1 4854e:1d MAY, 1986

ATTACHMENT 440.161

b. Low Head Pump Discharge Header Temperature (TE-912, 913, 914, and 915)

O There is one temperature element in the discharge header of each low head pump with readout on the main control board. These temperature trans-N mitters represent the inlet temperatures to each RHR heat exchanger and they are recorded, in conjuntion with the RHR heat exchanger outlet tem-perature {TE-924, 925, 926, and 927), by a dual-point recorder on the main control board to indicate the delta temperature reduction of the RHR flow.

b)

V These temperature elements are also used to provide input to temperature channel bistables that are part of the protection logic used to ensure that component cooling water flow is initiated to the corresponding RHR heat exchangers. The automatic opening of the RHR heat exchanger /

component cooling water isolation valves would be initiated in the event that actuation signals were generated by the RHR pump discharge temperature logic. A temperature actuation signal would be generated when l 1 a single temperature channel bistable receives a temperature signal from a corresponding temperature element, higher than a pre-determined O temperature setpoint.

c. RHR Heat Exchanger Outlet Temperature (TE-924, 925, 926, and 927)

There is a temperature element in the outlet of each RHR heat exchanger downstream of the flow bypass return. Readout is on the nmin control board. The temperature is recorded in conjunction with the RHR heat exchanger inlet temperature (TE-912, 913, 914, and 915) on the main control board by a dual-point recorder to indicate the delta temperature reduction of RHR flow. ,

d. RHR Heat Exchanger Outlet Temperature - Local (TI-916, 917, 918, and 919)

There is a temperature indicator in the outlet of each RHR heat exchanger. It provides a means of performance verification and heat balance when used in conjunction with the RHR heat exchanger inlet temperature indication of TE-912, 913, 914, 915.

O WAPWR-PSSS 6.3-42 AMENDMENT 1 4854e:ld MAY, 1986

~

ATTACHMENT 440.173 TABLE 1.3-1 (Sheet 1 of 2)

DESIGN COMPARISON O Parameter or Feature RESAR-SP/90 RESAR-414 RESAR-35 RESAR-41 Residual heat removal

1. Initiation pressure (psig) ~ 400 ~ 425 - 425 ~ 400 l1

2. Initiation / completion ~350/150 ~350/140 ~350/140 ~350/140 temperature (*F)

3. Component cooling water 105 105 95 105 design temperature (*F)

4. Cooldown time after ~ 16 ~ 16 ~ 16 ~8 initiation (hr)

5. Heat exchanger removal [ ](a,c) 46.5 37.4 39.4 l1 (106 Btu /hr) (4 provided)

Accumulators

1. Number 4 4 4 4

2. Operating pressure, normal 600 700 600 600 11 (psig)

O 3. Nominal operating water volume, each (ft 3)

[ ](a,c) 1,400 850 1,500 O

O -

O v WAPWR-PSSS 4854e:ld 1.3-2 AMENDMENT 1 MAY, 1986 i

ATTACHMENT 440.170 TABLE 6.3-2 (Sheet 2 of 6)

INTEGRATED SAFEGUARDS SYSTEM COMPONENT PARAMETERS Low Head Pumps lt Number 4 I Type Horizontal or Vertical O Runout Flow Rate, gpm (a,c)

Runout Head, ft Design Flow Rate, gpm Design Head, ft Shutoff Head, ft Miniflow, gpm Motor Capacity, bhp Speed, rpm Discharge Design Pressure, psig O Suction Design Pressure, psig Design Temperature, 'F NPSH Required, at pump suction (ft)

NPSH Available at pump suction (ft)

Seismic Category I Design Code ASME III, Class 2 O

II Based on a horizontal pump.

WAPWR-PSSS 6.3-49 AMENDMENT 1 4854e:ld MAY, 1986

ATTACHMENT 440.190 15.5 INCREASE IN REACTOR COOLANT INVENTORY Discussion and analysis of the following event is presented in this section:

1. Inadvertent operation of the emergency core cooling system during power operation This event, considered to be ANS Condition II, can result in an increase in reactor coolant inventory, if the reactor coolant pressure falls below the

( i shut-off head of the safety injection pumps following a reactor trip.

An increase in the reactor coolant inventory due to CVCS malfunctions will be discussed in the " Reactor Coolant System" module.

15.5.1 Inadvertent Operation of the Emergency Core Coolina System Durinq

~

Power Operation 15.5.1.1 Identification of Causes ard Accident Description O

Spurious emergency core cooling system (ECCS) operation at power could be caused by operator error or a false electrical actuation signal. A spurious signal may originate from any of the safety injection actuation channels as described in Section 7.3 of the "I&C and Electric Power" module.

Following the actuation signal, the high head safety injection pumps will start automatically. However, since the shutoff head of the safety injection pumps is less than the reactor coolant system (RCS) pressure, no flow will be delivered to the RCS. The passive accumulator injection system and the passive core reflood tanks also provide no flow at normal RCS pressure.

O A safety injection system (SIS) signal normally results in a reactor trip followed by a turbine trip. However, it cannot be assumed that any single fault that actuates the SIS will also produce a reactor trip. If a reactor trip is generated by the spurious SIS signal, the operator should determine b WAPWR-PSSS 15.5-1 AMENDMENT 1 4854e:ld MAY, 1986 l

ATTACHMENT 440.190 (Continued) if the spurious signal was transient or steady state in nature and if the safety injection signal must be blocked. Following reactor trip the safety injection system will not deliver any flow until the RCS pressure falls below the shutoff head of the safety injection pumps. For a spurious occurrence, the operator would stop the safety injection and maintain the plant in the hot standby condition. If the SIS actuation instrumentation must be repaired, future plant operation would be in accordance with the Technical Specifications.

O 15.5.1.2. Conclusions This transient is not analyzed since inadvertent operation of the emergency core cooling system without reactor trip does not result in the injection of any borated water since the shutoff head of the safety injection system is 1

well below normal operating pressures. If the spurious safety injection i signal produces protection system actuation, the incident simply results in a reactor trip.

O i

1 O

WAPWR-PSSS 15.5-2 AMENDMENT 1 4854e:1d MAY, 1986

ATTACHMENT 440.195 b) The amount of fuel element cladding that reacts chemically with water or steam does not f.tceed one percent of the total amount of Zircaloy in the fuel rods in the reactor. l1 c) The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. The localized cladding oxidation limits of 17 percent are not exceeded during or af ter quenching.

d) The core remains amenable to cooling during and af ter the break.

e) The core temperature is maintained at an acceptably low value and decay heat is removed for an extended period of time, as required by the longlived radioactivity remaining in the core.

These criteria were established to provide significant margin in ECCS performance following a LOCA. Reference 15.6.4-2 presents a recent study in the probability of occurrence of RCS pipe ruptures.

2 In all cases, small breaks (less than 1.0 ft ) yield results with more margin to the acceptance criteria limits than large break.

15.6.4.2 Secuence of Events and Systems Operations Should a major break occur, depressurization of the RCS results in a pressure decrease in the pressurizer. The reactor trip signal subsequently occurs when the pressurizer low pressure trip setpoint is reached. A safety injection actuaticn signal is generated when the appropriate setpoint is reached. These s countermeasures limit the consequences of the accident in two ways:

l a) Reactor trip and borated water injection complement void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.

O WAPWR-PSSS 4854e:1d 15.6-2 AMENDMENT 1 MAY, 1986

9 O  :

O

^

, m u

3 w

p L ~

O x ...

5 O

I

~~

~

4 O

4 QUALITY Figure 440.208-A. Quality in the Uphill Side of Steam Generator Tubes During a 3-Inch Break O WAPWR-PSSS AMENDMENT 1 MAY, 1986 4854e:ld i

1

4 o nos:

O 1

o ooot O I o cosi G

S E

.i I o ooo 1

O co oos w

N

, eo O

Ia E a

ie E a a QUALITY

' Figure 440.208-B. Quality in the Uphill Side of the Steam Generator Tubes During a 4.313 in. Break l WAPWR-PSSS AMENDMENT 1 4854e

1d MAY, 1986 l

e een O

i o ooet e mi 0

N o oooi P

O co m w

oe O

Ia E 4

i4 E 4

4 1

O Figure 440.208-C. Quality in the Uphill Side of the QUALITY Steam Generator During a 6-Inch Break AMENDMENT 1 WAPWR-PSSS MAY,1986 4854e:1d

, - - , - - - , - _ - _ _ _ . _ _ . - - - ,-r_ . . .._-- - _ , _ _ _ - - - . - - _ _ - _ . - , - . - - . _ . , _ , _ _ _ , . _ - - - . - - - , . - - . --- - - _ - _ - , _ _ _ . - - .

C'00Gi O

=

O 0*0002 0 0*00St i O E

I l

0*c0c O

l 00'005 I

.I D'O l

L 4

l I E i

E 4

E i

E 4

E i

E d

E s

l;  :

l El - essa le03 Jat A0 M01108 3 A08v 13 A31 Juna sin l

Figure 440.210-A. Core Mixture Height for a 6-Inch Break With One High Head S.I. Pump AMENDMENT 1 WAPWR-PSSS 4854e:1d MAY, 1986 l

k

. _ _ - _ . . . . . . . . . . - _ _ _ . . _ _ ... .. - . . _ _ . . _ . _ . - . . . . . . ~ . - - . _ . - - . - .

0*0012

. O t

0*0002 i

i i

. O 0*00GI

=

Y 0*0001 i

!O 00'001 I

i 4

i C*0 1

i 9 9 = a 1 8 i

8 c l $ $ $

j tviseoenstlee 5)e

! Figure 440.210-B. RCS Pressure for a 6-Inch Break With One j High Head S.I. Pump AMENDMENT 1 WAPWR-PSSS 4854e:ld MAY, 1986 i

i 1

i

. . _ , . _ . _ _ . . . _ . _. . _ . . . _ _ -._.._..m _ . _ _ - . . . _ . .

4 0*00$2 O

E_

0*0002 1

J

l i

O 0*0051 ,

b O

, l' 4

y 0*0001 f

l 00*005

/

. / '

/

O .

l o o e o e o o o o

- O, k k k k k E. O. E o

?? C R C R C 2 W o

! em 13 A 31 3erlastw v3n03mm00 i

Figure 440.210-C. Vessel Downcomer Mixture Level for a 6-Inch Break With One High Head S.I. Pump AMENDMENT 1  ;

WAPWR-PSSS 4854e:ld MAY, 1986

C'0052 l

l 5

I

~-

' C'0002 c cosi i 0 m~

i

! r i

C*0001 fi l

i 00'00G i

oo I

E E $ $ II E 8 E

d i i i d i i E on taa), senaria sitionssie.

4 Figure 440.210-0. Pressurizer Mixture Level for a 6-Inch Break With One High Head 5.1. Pump 1

WAPWR-PSSS AMENDMENT 1

! 4854e:ld MAY, 1986 1

=_ .. . - -. - . _ _ . . . _ . . . . -. _ - . _ . -

8'00G2 s

O 0*0002 O

O O'0058 "a

E E

C'0001 0

00'00%

i 0*C I

i b . o

$ E 4 4 .

4 4 i O esp 13A31 lentsin svla elden 1

Figure 440.210-E. Upper Head Mixture Level for a 6-Inch Break With One High Head S.I. Pump

! ,) MAPWR-PSSS AMENDMENT 1 I 4854e:1d MAY, 1986 i

e oost

O "

5 o ooot O

O o oost O

E 7

o oooi lO 00 00G oo i iT i R

i 2

i2 .

e F

,O im 1931 w. ::= nossoio) en mis Figure 440.210-F. Mixture Level in the Downhill (Cold Leg)

Side of the Steam Generator Tubes for a 6-Inch Break With One High Head S.I. Pump s WAPWR-PSSS AMENDMENT 1 4854e:ld MAY 1986

.________.__.._.___m . _ . _ .. . . _ _ _ . . _ . _ _ _ . _ _ .-._...__._.___.m. _ . _ _ .__ ___ . . .

D'00G2 i

4

O '

3 C'0002 O 1 l

1 O 0*0051 0

1 E

7 0'0001 O

l co cos 1

o*0 1

o 8 8 c

8 2

i c

i

~

ia .

s iso 11 Ali lent In 1935 e001 Figure 440.210-G. Loop Seal Mixture Level for a 6-Inch Break With One High Head S.I. Pump i

WAPWR-PSSS AMENDMENT 1 4854e:ld MAY, 1986

(a,C)

.i l

{

l l

i

!(

4 1

l t

i I

i

.1 l

J l

i 1

1 l

1 i

i i A I I sC HM if. ! A l. , l' 1 1 s i

l hi hb ' 11. 5

• 0:tivt tJEl' UppC' L3iar0"16 L' e i l g r.

{

U C W ". C O ' C " !!)Ott 4

4 i

H

• P

[ F#6N, , (,. h+ IeI lD w .E h hsi k

s Mar. 19Ct

.---=_ . - . . - . . . . - . . - - - . . . - -

4 I

t e

i ir l (a,c) 1, i

e i

i i

l 2

4 1

1

!e i

4 i

t I a

f

}

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e

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}

l l

l ,

i O i i

I i

l I

i O

A l 1_A_C _h..f1. .E.N..: . _4.i..l_1.

.. 4.3 ..B ,

AFWF (15:,5'c active tueli 9 Opper talencria L'esicr hencoter layout i

e I

l P

l -

WO We-F5:: AMENDMENT 1 nay, 1486 i I

(a,c) 1, I

l l

I r

I I

l l

l l

AIlkCHMtih, 4 4 0, 211_.C APWD t 15.1. 5 t.' HCt1VE +UA1' l

6 Uppe" CaisrC"le U0510' i.! Nc::le ' spetirer 6u1CE I

i Q- Q.C N May, 148t

(a,c) i i

i t

l l

t l

I t

ATTA H h t' N i 440,211-D A F M- t153,5i> H:tive tueli Upper Calandr:a be51gr.

5,I ho::le f Specimen bul0E AMENDMEt,1 1 WAPwc-FE:s

~

May, 196e i

000'04 O ~

000*St a

O f 000'52 O

E 000'025 ~

/

000'51 O

/ oo0'cx 0000's o / o'o 9 9 *

• 8 8 8

• 9 9 sie i e i e i e  :.

35n5$15d Figure 440.215-A. Upper Plenum Pressure Versus Time DECLG CD = .8 AMENDMENT 1 MAPWR-PSSS MAY, 1986 4854e:1d i

000'0*

O 000'st l

O I 000*0t o

I O I 000's?

O E

000'025 -

/

000'51 O

] 000 ci 1

0000's 7

O 1 c*o

• 9 9 9 9

• 9 g g g O k $  ! $

1snssive

! N N N I

Figure 440.215-B. Downcomer Pressure Versus Time,DECLG CD = .8 i

O Nodes 11,12,13,14 s

AMEN 0 MENT-1 MAPWR-PSSS MAY, 1986 4854e:1d

000*09 O '

000*GC 3

1 'l h 000*0t

. O # 0,,.,,

e 00001i ~

/

000*$1 O

/ 000*01 0000 s r

/

O / 00 l O

• g 9 * = =

9 8 8 l

E -

RE

~~

R E E E R E .

3trl5$3ud Figure 440.215-C. Downcomer Pressure Versus Time DECLG CD = .8 Nodes 47, 48, 49, 50 l

WAPWR-PSSS AMENDMENT 1 4854e:ld MAY, 1986

000'0*

O fr 000*St 000'Ot

[ '

l f '

O 4 l

l i'

l' h_

a 0000zE -

)

000*$1 l

000 0 s

G mw 0000 5 0*0 l

l

$ Y I O i i i I Y E N s M = 4 4 E 4 a a A ~ = y ? i T 3tvun01551 31vtM013-2 Figure 440.215-0. Downcomer Flow Versus Time DECLG CD = .8 Flow Links 11, 12, 13, 14 O

WAPWR-PSSS AMENDMENT 1 4854e:1d MAY,1986 l

00o*04 O

\

v \\ oco*St

\ >

i b  !

s coo oc J

. o N k oco sz

(\

(I.

C E

coo'otE

• dl l

L oco Si O ,

1. " oco ot 4

W =-

e c"TC >

= ~ -,x ,,en.,

o < x > oo I

l l

O

• n n

i i

31veNoI551 31vum01J-r Figure 440.215-E. Downcomer Flow Versus Time DECLG CD = .8 Flow Links 47, 48, 49, 50 l

WAPWR-PSSS AMENDMENT 1 l i854e:1d MAY, 1986

000*00 O ' 00o st 2

3 t

O x aco ot o

m Ooo se G

E Oooaz{

1 coo si coo os ecoa s O oo O i I

a I

E E 2 31 veno 1J !

Figure 440.215-F. CRT Flow Versus Time (2 CRT's Delivering to Downcomer Node 12)

WAPWR-PSSS AMENDMENT 1 4854e:ld MAY, 1986

ooo o.

O oco se

?

a O

ooo oc N w N

O w ooo se O

=

00002E -

oco si coo on oooo s i

O oo

.i i

I I I E E 2 317tmo1J-2

. Figure 440.215-G. CRT Flow Versus Time (2 CRT's Delivering to Downcomer Node 48)

AMENDMENT 1 WAPWR-PSSS MAY, 1986 4854e:1d

, - - , . - = . _ _ _ _ _ , , - . - - _ _ _ . - . - . . - , - - - - - - -

l 1

l O i 1

i lO

.8- ~

.7--

O E i

p .6--  !

o x 5--

w

$ .4 "

i S "3- -

a U

'O .2--

l

.1--

l 0.0  ; l O.0 50. 100. 150. 200. 250.

! TIME (SEC) l O

i O ,

Figure 440.215-H. Liquid Carryover Fraction Versus Time DECLG CD = 0.80 l O WAPWR-PSSS AMENDMENT 1 MAY, 1986 4854e:1d

Figure 440.218-A (Sheet 1) (a,c)

SATAN Input 9 - ,

l O

O f

O (a,c) 1 AMENDMENT 1 MAY, 1986 l

WAPWR-PSSS

Figure 440.218-A (Sheet 2) (a,c)

SATAN Input -

l I l 1 I i l 1 l

I l

l

, l l l l

I i l l l

-- (a,c) '

{

l O AMENCHENT 1 MAY, 1986 Iummunet WAPWR-PSSS

Figure 440.218-A (Sheet 3)

p. SATAN Input (a,c)

F e

1 9

9 1

l

C i J-O O WAPWR-PSSS 4854e:ld AMEN 0 MENT 1 MAY, 1986

- (a c) l I I l I )

I i l

l l

)

\

O O _

'~

FIGURE 440.218 SATAN NODING SCHEME FOR LARGE BREAK ANALYSIS AMENDMENT 1 WAPWR-PSSS (SHEET 4) MAY, 1986 4854e:1d

i

__ (a,c) 1 i

i t

i l

J 4

x 3

! n.

x o

h.

i I '

w a

E

=c m

i

<s u.

a m.

I e

i ko -

1 w

e l L.

3 m

u.

t i

i l

l _ _

Figure 440.218-B. WFLASH Model for APWR

) WAPWR-PSSS AMEN 0 MENT 1 MAY, 1986 4

4854e:ld

I TABLE 440.218-B DIMENSIONAL DATA FOR APWR WFLASH MODEL coNinot votunts _ (a,c) i I.

I

- _ (a,C) i i

l

)

.i I -

MODIFICATION OF UPSTR[AM AND DOWNSTREAM ELEVATION FOR Flow PATH (a,c) l AMENDMENT 1 WAPWR-PSSS MAY,1986 4854e:1d

< ATTACHMENT 440.220 The results show that the large break LOCA transient is characterized by three l

distinct phases: (1) initial heat-up phase due to the core stored energy and )

the relatively poor positive core flow period leading to the initial clad temperature rise (2) blowdown cooling phase characterized by influx of water I into the core from the large upper plenum and upper head volumes resulting in a significant blowdown cooling ef fect, and (3) reflood phase with relatively high core velocities driven by water elevation head in vessel downcomer.

' The maximum clad temperature calculated for a large break is [ ] which is (d less than the acceptance criteria limit of 2200*F of 10CFR 50.46. The maximum local metal water reaction is less than 1.0 percent which is well below the embrittlement limit of 17 percent as required by 10CFR 50.46. The total core metal-water reaction is less than 0.3 percent for all breaks, as compared with the 1 percent criterion of 10 CRF 50.46, and the clad temperature transient is terminated at a time when the core geometry is still amenable to cooling. No fuel rod burst occurs. As a result, the core temperature will continue to l1 drop and the ability to remove decay heat generated in the fuel for an extended period of time will be provided.

Small Break Results As noted previously, the calculated peak clad temperature resulting f rom a small break LOCA is less than that calculated for a large break. The limiting small break was found to be less than a 10 in, diameter rupture of the RCS

< cold leg. A range of small break analyses are presented which establishes the limiting small break. The results of these analyses are sumarized in Tables 15.6.4-4 and 15.6.4-5.

Figures 15.6.4-31 through 15.6.4-39 present the principal parameters of 1 interest for the small break ECCS analyses. For all cases analyzed the following transient parameters are presented:

a) RCS pressure b) Core mixture height,

(

O WAPWR-PSSS 15.6-12 AMENDMENT 1 MAY, 1986 1 4854e:ld

2 O <

ac 2

Eu 5

5E by ..g e-oc m

=

O

. .oc I S G

a E

S w -

5 i r

i

! --8 m I

i 1

!O  :  :

l o '

e o o e o o o

! R 8 8  ? 8 8 2 (W97 30 SONVS00H1) AB01N3ANI i

i i Figure 440.221-A. RCS Inventory for a 3-Inch Break i

WAPWR-PSSS AMENDHENT 1 4854e:1d MAY, 1986 4

i

! O O O O O O O i are cn >

t a

• r

?

.- ,?

"$" 2 35- VESSEL COLLAPSED LIQUID l $ LEVEL FOR A 3 INCH BREAK P

~ -

". E 30-o  !

1 5" l 1 u.

t o T

i r i $.-

8 1

i

! 8 25-i 9 w

= b a ti j

r- g 20-t @ a

~

i C ""

~

.\ r- a e

% ~

15-TOP OF CORE (J

h n

10 , . .

• 0 50 1060 1500

?%

=

TIME (SECONDS) ,

1

-@.x 3

w CD Z 1 e- ,

_. . . _ . .. _ . _ - . - _ . _ - - _ _ . - . .~ . . . _ _ . _ . _ . - . _ . . . . - - _ _ .__.. .. - - . - - . -

i c' cost  !

I l

I i

1

] 1 1 5 o ooot I "

- O'0058 U

e E

s, I

?

0*C301 3

1 l 1

i

/

1 00'005 i

t

! c'0 a

k k b $ $g $

• 2 R 4 g g . ,

11J8 13A11 lenisin Oy)M elden Figure 440.221-C. Upper Head Mixture Level for a 3-Inch Break i WAPWR-PSSS - AMENDMENT 1 4854e:1d MAY, 1986

0*0052 l

I i

o*000%

'i 1

4 i

O*0051 O

E E

i

! c ooot i

l -

- )

1 00'005 i

i l

0'0 i

4 t

4 4 c

4 2

4 2

i a

i s

R 4

g lin 11A11 lentalm 1935 4001 Figure 440.221-0. Loop Seal Mixture Level for a 3-Inch Break WAPWR-PSSS AMENDMENT 1 4854e:1d MAY, 1986

..n-.,- -,----

-~,,-,---_--n ,-,-- ,,_ ,_.n_,--- ,_ n-, ---n..-.-- , , . . . . , . . - _ _ . . -.-

_ . . . - - - . . - _ _ _ _ . . _ _ . . . - . - - - . -_ _ _ . - _ ____ _ = ,

3 0'00GF 5

0*0002 l

O 0'00$1 O

w t?

I E i

J 0'00]I I

1 i

i Co Oos

/

. - /

O .0 i

E E E E E i i i i si E in 13431 juntain saisoici nis nas i

i Figure 440.221-E. Mixture Level in the Downhill (Cold Leg) Side of the Steam Generator Tubes for a 3-Inch Break i

WAPWR-PSSS AMEN 0 MENT 1 4

5854e.1d MAY, 1986 l

4 o me i

< o ooor 3-j l o 005 i

' 0.,

t

! 7

=

! o*0001 4

i l

co cos I

6 I

1 '

s -

/~

, o*C l

i E E E $ E I E g

i i i i i d i  ;

-

im 13^31 isnuin airienssses Figure 440.221-F. Pressurizer Mixture Level for a 3-Inch Break AMENDMENT 1 WAPWR-PSSS 4854e:1d MAY, 1986 l

l n'00st I

i I

I 0*0002 I

, $l-h O'0051 O

M E

i f O*0005 1

O d

00'005 i

i O , 0'o i

kk 7?

I E

k E

C E

i E

ti E

s j

j y l

41913 A 31 Junt ain ajwo3nneo Figure 440.221 -G. Downcomer Mixture Level for a 3-Inch Break

O WAPWR-PSSS AMENDMENT 1 MAY, 1986 4854e:ld

O O ,

i arz co >

T2 ox i

" 4, i

a. m

$ 700 RCS INVENTORY FOR A 2

e 4.313 INCH BREAK C,

• 600 "

b o

o ^

r  !

ru.

- m p [ 500-- i o

= E z

'm' -t

~

= $ 400 --

o

< z E

• E x E o

o g 300 -

' w a 5 '

i l

b b3 200-- -

.L i =  !

, n '

I 3" I

%' 100  !

2 0 50'0 10'00 15'00 w t x> TIME (SECONDS)  ;

>1

<m  !

• Z O

&E

@m i CD Z  !

cn -4

*rz co >

m -o

=r i ox "

1

, ,4 -- 35 ,

l ng E VESSEL COLLAPSED LIQUID  ;

l u' 2 LEVEL FOR A 4.313 INCH BREAK

.as o.

m

~ m 30 --

i w

~

w ,

< w

= o I

• r a o l 25 --

4

! c, o i

o m

w

, o, >

u o w m o <

" e-r- w 20--

w i a

• 5 l i

i $ d t

> i

i. r.- w I

1 a

' 15--

J o

TOP 0F CORE De

.b I S 10 t l Y 0 500 1000 1500 1

E TIME (SECONDS) l 2,>x = ,

t

,h G3 jj _. a, 'e i

@m ca x n' I OS -4

=* i I

)

w w --r

I c oose 4

s

?

^

o ooot i

i i

i j

t o ooss I

O

=-

Y 1  :

l o occi 1

i 1

i i

t i

1 1

/ co cos i /

/

> ,/

l 1,

oo i

i 4

I I $ $ $ =

~

1 , e .

i sap 13A31 3entalpi Ovla 51 den i

!, Figure 440.221-J. Upper Head Mixture Level for a 4.313-Inch Break j

AMENDMENT 1 I WAPWR-PSSS MAY, 1986 i 4854e:1d 4

i l

1 , -- .- - . - .

.- --..-.c-,-,.,----- - - - - - - - - . . - , - . - - - . . - - - - - - . . - - - _ . . - - -

i 1

I o oou i

1 i

i i

~

! o ooor i

1 1

o ooit l 0

l 7 1

I r 5

1

! o occi t 1

I i I

i r

I co ocs

' (

1

, # l l

i l I

\, l 00 l

! l l l I

i N

i I

e A

a I

i J

ls la  :

l alp 13A31 lent:In 1938 eo01

!a

. Figure 440.221-K. Loop Seal Mixture Level for a 4.313-Inch Break i I AMEN 0 MENT 1 )

i WAPWR-PSSS 4854e:1d MAY,1986 i i

i 1

1

! 0*0052 i

i I

l 1

! I a i i  : +

i 0*0002 l e

4 i

i i ,

l 4 0*005: i 1; a"

,, E tes j i f, i

~

1 i >

I j 0*0001 1

'c, t

1 I

4 I

l i

' C0'004 I a

4 i

i f

i #

4 /

]

I n.n i

t E. E. I. I. E. .

1

,q T

• e a e eso innis ionssin joisoto) 'an was  !

Figure 440.221-L. Mixture Level in the Downhill (Cold Leg) Side of ,

the Steam Generator Tubes for a 4.313-Inch Break l i_  !

i l

AMEN 0 MENT 1 i i WAPWR-PSSS j 4854e:1d MAY, 1986 l f

l

. _ _ _ _ _ _ . _ . _ . _ - _ _ _ _ _ , _ _ _ . _ _ _ . _ _ _ . . . _ _ _ _ _ _ _ _ __o_

o oost O

i O o oco:

O o coGI O

E I

o coo O

co cos oo I I I ls O 1 I

e I

e I

e i e I

e tan il A31 lena sin elitentslee Figure 440.221-M. Pressurizer Mixture Level for a 4.313-Inch Break WAPWR-PSSS AMENDMENT 1 4854e:1d MAY, 1986

. _.. . . .. - - .- - . _ . ... ._ - - _ . . . . - . _ . . _ _ = - _ . - - _ .

I c oct2 fl i

i 3

-i 3

o 0002

)

I

i. '

- o octi 4

1

@ 0 i 3 1

3:.

T l

i i

i o oooi i

i i

I i J co cot 1

i

)

l1 co t

I 4

l 1

@E E E E E E E l

i s e e n i e e s  : '

esp 11 A 31 1 ens sin elat3mm00 Figure 440.221-N. Downcomer Mixture Level for a 4.313-Inch Break ,

t I

AMENDMENT 1

! WAPWR-PSSS MAY, 1986

! 4854e:1d l

.i 6

- - ,-.____.,-,,.___m_,_ _ __.__....... .... - , _ .,-m.,,,_ ,-.,, . _ . . . . _ _ _ . - _ _ ...-..,,-,,m.

. _ _ _ . - - - _ _ _ . - . _ _ . ~ . . . . . . - - _ - _ . _ - _ - . _ - - - _ _ - - - _ . _ . . _ .

il h \' ::

1 l -

l

s l' 1

, l

.; e 2w E6 o=

rm s

! --8 e

me o -

a<

O l

1 i o

. 8 n 4, '

5 v

kJ 4, m l

l !E ~

!O

~

8 1

i a

i I

lO j

. l l l l o l 8

~

8 e

8 m

8 e

8 m

8 m

8

(W97.40SONVSn0H1)AB01N3AN!

Figure 440.221-0. RCS Inventory for a 6-Inch Break AMENDMENT 1 WAPWR-PSSS i

4854e:1d MAY,1986 l

I O o E l

I Es om u i U 5 i S5 O

I .

E. l . .g 5 .c =

o E "2 l da [

l l

l I

l .

8 I "g l 8 w

I e.

I W O l I

I y 8 m

gi 5

al El 1

- 1 O C ;  :

l l

. l a E 8 N 8 $ S (380330WO11083A08V1333)13A37 O i i

1 Figure 440.221-P. Vessel Collapsed Liquid Level for a 6-Inch Break l

\

O WAPWR-PSSS 4854e:1d AMEN 0 MENT 1 MAY, 1986 l

i

O*0052 O

I h

O 0 0002 O

0*0051 O

M Y

0 occi O

00'005 0 00 k 5 k k 2 2 k k .

4 J .

4 g im 13 A313sna rin Ov 3M v3 den Figure 440.221-0. Upper Head Mixture Level for a 6-Inch Break

AMENDMENT 1 WAPWR-PSSS 4854e

1d MAY, 1986

't

,_ _ . . _ _ _ _ ~ . _ _ _ _ . _

. .- - - . = _ . .

. . . . - - . - - .. - _ . - . - - _ _ . - . _ . . . - = . _ . . _ _ .

P

}

c'0052 O

I 5

0 0002 c'0051 0

h-E

=

'\

0*0001 O

l 00'00%

I

^

O F 0., i l

o 2 O C 2 R m J 4 d tan 1]A31 3snisin 1935 4001 4

Figure 440.221-R. Loop Seal Mixture Level for a 6-Inch Break

' O WAPWR-PSSS AMENDMENT 1 4854e:1d MAY, 1986 3

,ve--,-. y . ,-..w- - - . - y----

l l

0*0052 O l 3

1 3

0*0002 O 0*005I 4

b d

l Y

• \

l 0*0001 O i 00'005 ,

i O ,

0*0 0 8 i

?

e R

i (AD 13A31 3 nt IM 30l$0103 *N33

• Mis i

f y

- Figure 440.221-S. Mixture Level in the Downhill (Cold Leg) Side of

( the Steam Generator Tubes for a 6-Inch Break WAPWR-PSSS AMENDMENT 1 4854e:1d MAY,1986

0'0052 O

I 0*0002 O l O*0051 i

O E

E C

0'0001 O

00'005 0 0*O f

b $ E $

)

I E A C N d $ y tan 13A313 nt In g3rlan551sd i

Figure 44 0. 221 -T . Pressurizer Mixture Level for a 6-Inch Break WAPWR-PSSS AMENDMENT 1 MAY, 1986 4854e:1d

00WNCOMER MI Tunt (tytt art o

• *  ? 3 3 5 3 i s I

l 'I I I I I 'I 3  ;

l l

0.0 l O I O 500.00 /

\

1000.0 4

l A

3 1500.0 O 2000.0 ,

5 O

2500.0 i Figure 440.221-U. Downcomer Mixture Level for a 6-Inch Break WAPWR-PSSS AMENDMENT 1 4854e:ld MAY, 1986 l

O ATTACHMENT 440.225 TABLE 15.6.4-1 O INPUT PARAMETERS USED IN THE ECCS ANALYSIS 0 (a,c)

O Core Power

• Peak linear power (includes 102% factor Total peaking factor, F g

Power shape Large break-chopped cosine Small break-see Figure 15.6.4-39 Full assembly array Accumulator water volume (minimum)

I Accumulator tank volume Accumulator gas pressure (minimum)

Core reflood tank water volume (minimum) 11 l Core reflood tank volume Core reflood tank gas pressure (minimum) I1 Safety injection pumped flow j Initial loop flow Vessel inlet temperature j vessel outlet temperature Reactor coolant pressure Steam pressure O Steam generator tube plugging level -

l l

O

• 2% is added to this power level to account for calorimetric error.

O WAPWR-PSSS 15.6-21 AMENDMENT 1 2854e: MAY, 1986

O O O O O O O l

l s' l  ::E

! "aT",

u i v>

100 m 8 Z g 6 c

4 -

m 6 2 >

( TOTAL RESIDUAL HEAT d d,

• 10-1 : iE; (WITH 4% SHUTDOW) l8 6

~

J. I 9 '

- H S' t g 4 -

2 i

5 32

e g

g 10-2 8 =

6 Z g 4 -

g -

• ~

k

1. 10-3 I I IIIIll l I IIlill l l l llllll l l l llllll l l ll lj;i 10 1 2 5 100 2 5 101 2 5 102 2 5 102 2 5 104 EE

" Time After Shutdown (sec)

'5

. "i!!

E 'i 4

(a,c)

O:s M

E-t/]

CO CD Q

05o O

M k

j 4 g

Q M

P Oso.

M E-Z I

5 O%

O O Figure 440.239A. ISS Flow Diagram AMENDMENT 1 WAPWR-PSSS MAY, 1986 4854e:1d

O O O O O O O

c5 T2

• ?

EBm APWR - CHEMICAL & VOLUME CONTROL SYSTEM i

i a

T 1

2

~

W

=

O, c

! M i

T o

c a

a a

3 i

??

-E 29 em a

S.

-