NRC-89-3438, Forwards Nonproprietary Draft Responses to Open Issues 42-81 of Draft SER Re RESAR-SP/90 Preliminary Design Application

Forwards Nonproprietary Draft Responses to Open Issues 42-81 of Draft SER Re RESAR-SP/90 Preliminary Design Application

ML20245B085
Person / Time
Site:	05000601
Issue date:	06/09/1989
From:	Johnson W WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NS-NRC-89-3438, NUDOCS 8906230009
Download: ML20245B085 (43)

Text

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b Westinghouse PowerSystems Nuclear Technology SYSter"S D*Slon Electric Corporation Box 355 Pittsbutgh Pennsylvania 15230 0355 June 9, 1989 NS-NRC-89-3438 Docket No. STN 50-601 Document Control Desk U. S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Charles L. Miller, Director Standardization and Non-Power Reactor Project Directorate

Subject:

Submittal of Westinghouse Response to Draft Safety Evaluation Report (DSER) Open Issues 42-31 in Review of.RESAR-SP/90 PDA. j

Reference:

C. L.. Miller (NRC) letter to W. J. Johnson (Westinghouse), dated March 9, 1989.

Dear Mr. Miller:

Enclosed.are ten ~(10) non-Proprietary copies of the Westinghouse draft-

. responses to Open Issues 42 through 81 of the staff's Draft Safety Evaluation Report-(DSER) (Reference) in their review of the RESAR-SP/90 Preliminary Design l Application.- The responses include the text changes that will be made in our  !

final submittal and commitments to meet the requirements of some open issues at a later stage in the licensing process. Responses to DSER Open Issues 1 through 41 and 82 through 107 will be forthcoming in the next two weeks. Some of those open issues are currently being discussed with staff members in order to obtain. agreement toward achieving complete closecut.

Note that' the enclosed package is agi to be inserted in the RESAR-SP/90 l licensing document. The formal submittal will include the required number of 1 Propriett.ry Class 2 and non-Proprietary copies of all text changes and responses for inclusion in RESAR-SP/90 PDA licensing document. 1 Very truly yours, 8906230009 890609 3

-PDR. ADOCK 050 1

[ -U Qu J. Johnson, Manager

-/c%JI u Nuclear Safety Department Enclosure (s) l .WMS/bek/0434B cc: Thomas J. Kenyon - NRC (MS 11H3)

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R Response.to. Draft Sa_fety Evaluation ,

• ' Report;0 pen' Issues 42 .

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'DSER Open Issue 42:- Integrated N16 and excore power density survelliance and-protection: system (4.3.1, 4.3.3, 15.3.2).

Response

This open issue involves a continuing review on the part of the staff, and Westinghouse will_ take no action unless the results of this review are issued prior.to +he Final-SER.

'DSER'Open Issue 43: Review of critical heat flux (CHF) tests (4.4.2.2).

Response

.( 'This open issue involves a continuing review on the part of the staff, and Westinghouse will take. no' action unless the results of this review are issued

"- prior to the Final SER.

DSER Open Issue 44: Departure from nuclear boiling ratio (DNBR) safety limit (4.4.3.1).

R_esponse:

This open issue, which is contingent upon the resolution of Open Issue 43, involves a continuing review on the partfo' the staff, and Westinghouse will take 'no action unless the results of this review are issued prior to the Final SER.

DSER Open Issue 45: ASME code case commitments for all ASME Class 1, 2 and 3

, components (5.2.1.2).

. Response: i

-Amendment 1 (May 1986) to RESAR-SP/90 PDA Module 7, " Structural / Equipment L Design," in response to Question 210.20, does provide the required commitment for components other than RCS components, although it doesn't 'nention WAPWR-SER 1 MAY 1989

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!Clisses 1,. 2 or 3~ specifically. To help . ensure that this commitment is applied as; intended, the first two lines of the second ' paragraph on page i 210-11 of Amendment 1 of.RESAR-SP/90 PDA Module'7 will be revised as follows:

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"When conditionally approved, Code Cases are used in the construction of ASME Section.III Class 1, 2, 3, MC, NF and NG components, the additional conditions: identified in Regulatory Guides 1.84 (Design and Fabrication),

1.85 (Materials).and 1.147 (Inservice Inspection) as applicable to these conditionally approved Code Cases will be complied with."

DSER Open Issue 46: Pressurizer safety valve sizing (5.2.2.1).

Response

To make the required commitment, the third sentence in the first paragraph on page 5.2-4 of RESAR-SP/90 PDA Module 4, " Reactor Coolant System," has been.  !

revised as follows:

"The reacter is maintained at full power, with no credit being.taken for either the reactor trip on turbine trip or the first safety grade reactor trip.  : Credit is taken for steam relief through the steam generator safety valves."

DSER Open Issue l

2 4: Low-temperature overpressure protection'(LTOP) . during l plant startup (5.2.2.2).

m Response:

The use of the RHR relief valves to provide low temperature overpressurization protection (LTOP) was evaluated for both plant startup and plant cocidown operations. The current statement in Subsection 5.2.2.10 of RESAR-SP/90 PDA Module 4, " Reactor Coolant Systems," that two of the four ISS RHR subsystems be aligned to provide acceptable LTOP protection, is consistent with startup ,

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x operation requirements.. .Specifically - for startup the evaluation discussed in LSubsection 5.2.2.10 of Module'4 included:

o- A low Appendix G pressure limit for the RCS for startup/heatup.

t V o .An analysis of the APWR heatup operation to establish RCS temperature vs.' pressurizer. pressure throughout the startup operation.

This' evaluation 'shows that RCS heatup to ~350'F is accomplished without challenging the RHR relief valve set pressure and that LTOP protection is not required above.~350*F.

DSER Open Issue 48: LTOP during single failere of residual heat removal (RHR)

-valve (5.2.2.2).

Response

The Westinghouse commitment to remove power from the open RHR suction

. isolation valves ensures that no single failure (i.e., single inadvertent operator action or energization) can cause any of these valves to inadvertently close and result in an RHR relief valve being unavailable for overpressure protection. Also, as discussed in the response to DSER open issue 49 below, an alarm will be provided to alert the operator when LTOP protection is required, to assure that at least two RHR subsystems are aligned, and that all four valves in at least two subsystems remain open.

Furthermore, plant procedures will state that maintenance during plant shutdown should be limited to one of four RHR subsystems when overpressure protection is required. 1 J DSER Open Issue 4_9: Positive indication /elarm to signal need for initiation of LTOP system (5.2.2.2).

Response

Westinghouse will comply with the requirements of BIP RSB 5-2 for LTOP instrumentation as outlined belom WAPWR-SER 3 MAY 1989 B197e:1d

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o An alarm wil'1 be'provided to alert the operator to enable LTOP at the i correct plant condition during cooldown. The alarm will ensure that protective action is initiated and maintained by requiring that both isolation' valves in two of four RHR subsystem suction lines are open. f o Positive indication that LTOP is enabled is provided by the position indication IQits for the RHR suction isolation valves.

DSER Open- Issue 50: Emergency feedwater storage tank compliance with Position G of BTP RSB 5-1 (5.4.3.1). i i

Response

Position G of BTP RSB 5-1 is complied with in the RESAR-SP/90 design. The !

required commitment is included by revising Subsection 10.4.9.2.1.2 on page  ;

10.4-20 of RESAR-SP/90 PDA Module 6/8 " Secondary Side Safeguards System / Steam and Power Conversion System" to read as'follows:

"Two equally sized emergency feedwater storage tanks are provided, one in l each subsystem. The tanks are safety grade, seismically qualified and E protected from missiles, fire, etc. The tanks contain a sufficient quantity of' condensate quality water to permit operation at hot standby i

conditions for eight hours, followed by cooldown to the conditions  !

permitting operation of the residual heat removal subsystem. The required emergency feedwater inventory in the tanks is based on both of the following:

1. An extended time (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) to cool down to RHR cut-in conditions is assumed, based on a natural circulation cooldown associated with rnly onsite power being available. This conservative assumption maximizes the decay heat input to the RCS during the cooldown.

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j y - 2. One reactor coolant: pump is--assumed to. be operating during the

R cooidown) continuously adding heat to the RCS. This is consistent with .only; onsite power being available, and is. an additional conservatism--which maximizes :the total heat input to-the RCS during ~

the,cooldown.

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k In~ addition, the combined water inventory of both tanks is sufficient to -

f allow' indefinite . hot' standby operation. An indefinite time is satisfied by:

,1 a. having one day's supply of emergency feedwater stored in the EFWST's L. and taking- credit for availability of the alternate water supplies in

( 'that time, or.

b. use of primary side feed and bleed operations.

Both the cooldown and indefinite hot standby operations can be conducted l using only- safety grade equipment and with the most limiting single failure assumed. i

- The -tank design .provides- additional volumes of water to allow' for

- inaccuracies' in the tank level. indicators. overflow and pumpout margins and spill allowance."

DSER Ooen issue 51: Decay heat generation rates (5.4.3.1).

Response

The following was provided in Amendment la to RESAR SP/90 PDA Module 1,

" Primary Side Safeguards System," to amend our original response to.NRC Question 440.41.

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• generation rates .which would occur, and.thus provide an acceptable

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Lbasis'for. sizing decay heat removal. system equipment. +

Q, m  ; .g b, The decay. heat' generation rates in ANS Standard 5.1 do not contain 1

the conservatism, which have; historica11y' been employed by the-fh.

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@, ' NRC.; For example, the decay heat generation' rates in BTP ASP 9.2 4 linblude a _20% -uncertainty. for 'the first '103 seconds followin reactor ishutdown, and a 10% uncertainty;between 103Land 10

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' seconds' after- shutdown. Such conservatism may have been ,

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considered necessary.in the past to t envelope actual decay heat

, g .' removal pates, but are . unrealistically high. based on current

- knowledge. ~In.~particular, to quote from the foreword of ANS 5.1-1979:

4 4 "In 1974,- .new research programs were initiated under the auspices w

yT . of the Energy ,'Research and Development Administration, Nuclear

>v . Regulatory Commission, and Electric Power Research Institute to h hf " better quantify decay heat and its uncertainty for short cooling-

)y@ , . times. The -ANS 5.1 Working Group- was reconstituted to include l~

those individuals engaged in the new research and representatives

'( from industry and NRC who have knowledge of decay heat from those

- perspectives. The first objective of the Working Group was

.; defined; to be a revision of the ANS 5.1 standard for LOCA 4

> N. applications (cooling time up to 10 seconds) in LWRs. The 1 present- a revision 'provides precise results, including detailed evaluation of the influence of neutron capture in fission products

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[. -for this shutdown time range. It also covers the cooling times up to 109seconds by use of an upper bound for the capture effect."

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.l Thus, the intent of ANS 5.1-1979 was to incorporate- the best available. I knowledge in defining decay heat generation rates suitable for LOCA analyses.

Subsequent evaluations have shown ANS 5.1-1979 to be conservative, compared to the most realistic evaluations of decay heat generation. One widely accepted code.for realistic determinations of decay heat generation rates is the ORIGEN* _ code. A recent comparison ** of ANS 5.1 (1979) results to ORIGEN results showed that ANS 5.1(1979) results were conservative by about 3 to 5%

3 for the first 10 seconds following shutdown, and by as much as 18% in the 3 6 range of 10 to 10 seconds.

, Accordingly,'ANS 5.1 (1979) is considered an acceptable basis for design of SP/90 decay heat removal equipment.

DSER Open Issue 52: Power supply restorat' ion of motor-operated valves (MOVs) in RHR system return line from control room (5.4.3.2).

psponse:

The normally closed motor operated valve (MOV) outside containment in the RHR e

pump discharge flowpath will not have power removed. Therefore, operat.-

i action to open the RHR flowpath is accomplished by normal operator action in

-the main control room, as required by RSB 5-1.

• See, for example, Bennet, P. E., Sandia-ORIGEN User's Manual, NUREG/CR-0987, SAND 79-0299, October 1979.

• • Memorandum from F. Eric Haskin on "Whole-Core Decay Heat Power," dated January 28, 1986.

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p , DSER Open, Issue 53: Thermal relief protection for RHR (5.4.3.3).

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Response;:

Subsection 5.4.7.1 of RESAR-SP/90. PDA Module 1, " Primary Side Safeguards

,f System," has been revised as'follows:

" Thermal relief protection is provided between pairs of normally closed RHRS. isolation valves, as follows:

1. In the RHR pump suction lines from the RCS - Between 8 inch motor operated valves 9000A and 9001A, for example, by means of 3/4 inch check valve 9019A, which will relieve pressure from between valves 9000A and 9001A to the RCS side of 9000A.

2. In the'ISS pump full. flow test lines to the EWST -

Between 6 inch motor operated valves 8813A and 8814A, for example, by means of 3/4 inch check valvo' 8815A. This will relieve pressure from between ivalves 8813A and' 8814A to the upstream side of 8814A, which in turn is protected against overpressure by relief valve 9020A.

y 3.: In the connections between the RHR/CS pump discharges and the containment spray headers - Between 6 inch motor operated valves 9009A and 9011A, for exampii, by means of 3/4 inch check valve 9010A, which will relieve pressure from between valves 9009A and 9011A to the

1. upstream side of 9009A, which in turn is- protected against overpressure by relief valve 9022A. Note that 6 inch valve 9009A is normally open (aligned for containment spray); it is closed for RHR' operation."

l f DSER Open Issue 54: Lowered reactor coolant system (RCS) inventory operation of RHR (Generic Letter 88-17) (5.4.3.4).

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! Response: '

- Subsection 5;4.7 of RESAR-SP/90 PDA Module 1, " Primary Side. Safeguards System"

^ has been revised as follows:

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. "During certain; shutdown periods, it may be. necessary- to perform' inspection and/or' maintenance operations on the steam generators and reactor coolant pumps. Toward the end of the associated cooldown the LD reactor coolant inventory is' reduced sufficiently to drain the steam H generator channel heads and install steam generator isolation devices (nozzle dams). The RCS water. level is lowered while RHR operation is h continued; this~is termed."mid-loop" operation.

Following nozzle dam installation, the RCS water level is raised to the appropriate level for continuation of the inspection / maintenance work (just below the vessel flange) or for refueling (top of refueling canal), i unless the reactor coolant pump shaft is to be removed. Pump shaft removal requires that mid-loop operation be continued.

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.To ensure its' continued availability to perform the residual heat removal l 1

. function during mid-loop operation, the following features are  ;

incorporated in the' design of the reactor coolant system (RCS) and the L residual heat removal (RHR) portion of the integrated safeguards system 1

(ISS): 1

1. The layout of the RCS hot leg piping and the steam generator channel I i

head .is such that installation of the nozzle dams can be performed with an 80% level in the hot leg piping; this is 9.3 inches above the actual mid plane elevation.

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2. With the conventional Westinghouse arrang6nent of a residual heat j removal piping connection at 45* from horizontal, it has been )

calculated -that onset of vortexing with attendant air ingestion would l occur at a level 3.0 inches below mid plane elevation. Therefore, during "mid-loop" operation, a margin in excess of 12 inches would .

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exist betwe'en normal: operating level and the critical level at which-

RHR pumpsoperation may be impaired due to high levels of air entrainment. -W'ile h this- is a:.significant. improvement relative to current-plants, Westinghouse commits to install,_in addition, a vortex 1 breaker in each RHR suction nozzle.' This vortex breaker consists 'of a 24 ; inch long section of.14 . inch Schedule 140 piping connected in'a vertical direction to the bottom of the hot' leg piping; the 8 inch RHR c

suction line- is connected to the bottom of this vortex breaker. With

!a vortex breaker, air ingestion-' commences at about the same water-

. elevation- as 'with a conventional RHR ~ suetion _ nozzle; however, the

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amount of air entrainment will remain below 10% unless the hot leg is completely.. drained.- Therefore the potential for RHR pump' damage has essentially been eliminated.

3.- The RHR. pump suction line is "self.-venting," i.e., it slopes continuously upward' from the pump to its connection to the hot leg (vortex: breaker). If the pump should stop during mid-loop operation (due 'to interruption of electric power, for example) any air bubbles present in the pump or suction piping will be vented back up through the suction' line. to the water surface in the hot leg. This feature provides_for re-starting the pump under conditions which automatically

-assure a flooded suction.

4.' Separate . narrow range ~1evel transmitters, calibrated for low l temperature conditions, indicate the RCS water level between- the bottoms of. two hot legs and the tops of the steam generator inlet elbows in the same loops during the approach to and conduct of mid-loop operation. Indication in the main control room and low level alarms are provided.

5. The range of the wide range pressurizer level instrumentation used during " cold" operations, has been expanded to the bottom of the hot legs. This provides a continuous levei indication in the main control room, transitioning to the range of the two, more accurate, narrow range loop level instruments.

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cavitlstionl or L otherladverse effects under conditions of no subc' ooling

,in!th6jhot legs. Specifically, definition of design; values 1for_ "NPSH-

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Davailable," "NPSH ' required" (by the pump) and .the required layout a

!, characteristics (elevation; difference, pipei routing,: etc.)fwill be-coordinated' to' assure--that:~the RHR pumps can be started and run at-1their full RHR flowrate, even if boiling in 'the reactor -vessel is. 1 occurring. 'This Jassures that the normal RHR function.can befreadily $

used to. recover from a temporary loss of cooling.

7.a A: locally mounted' flow- transmitter in each RHR return header.

.(downstream cf :the RHR heat exchanger), with readout in the main icontrol. room, indicates RHR return flow to the reactor vessel. A low alarm will. alert the operator to a decrease in RHR flow in the associated subsystem.

8.. The, drain 1 down of the RCS to mid-loop operation. level and RCS

. inventory . control- during mid-loop operation is performed by the operator in'the~ main control room, using the RHR to CVCS letdown 0

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(flowpath. and inormal CVCS functions. This will eliminate the need to coordinate-local actions in the containment with the control room c operators to. control RCS drain down rate and level.

9; Procedures' will, require that. one of the_ four HHS! pump subsystems

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'always will be avai1able for use 'during mid-loop operations. This lwill ensure that a' backup source of water for restoring RCS inventory is readily available and can be actuated.from the main control room.

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10. At -least two incore thermocouple will' be available to directly measure the core exit temperature during mid-loop RHR operation. Each of these thermocouple .will be on separate instrument electrical

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channels. Also, 'since the SP/90 incore thermocouple are independent of the RV head, their availability can be maximized; howaver, these thermocouple- will be retracted from the core region during the actual movement / replacement of the fuel. It should be noted that when fuel is being moved the refueling cavity would be flooded.

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T, Note- that. these design features provide the operator ' in the main control room with..all. required instrumentation, alarms, and operation controls 9 s ,

necessary-to adjust, maintain, and take any necessary recovery actions for both RCS inventory control and heat removal.-  !

Additionally, during the Final . Design- Application (FDA) phase,

. Wwestinghouse.will. perform evaluations to examine potential design-' criteria

, to' establish procedures and Administration controls that will reasonably 1

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ensure that' containment closure will be achieved prior to the time at  !

which a core- uncovery .could result from _a loss.of RHR coupled with'an inability to initiate alternate cooling or addition of water to the RCS j inventory.- '

In. addition- to these design features, appropriate operating and emergency procedures.will'be defined to guide and direct the operator in the proper conduct. of mid-loop operation,- and to aid in detection and' correction of off-normal conditions which might occur during such operations."-

DSER'Open Issue'55: Boron mixing / natural circulation test (5.4.3.5).

Response

1 The Westinghouse response to Item 21 of Subsection 6.4, "NRC Generic Letters,"

of: RESAR-SP/90 PDA Module 2, " Regulatory Conformance," has been modified as follows:

"As part of the pre operational and initial startup test program, a combined natural circulation and boron mixing test will be performed.

This test will. demonstrate the capability of the RESAR-SP/90 design to

1. provide natural circulation cooling of the RCS, and ,

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'2. mix injected borated water in the RCS under natural circulation l conditions.

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The' resultsi of ' these tests,: together with' appropriate analysis, 'will-demonstrate Lthe capability' of 'th'e SP/90 plant to achieve a cold shutdown a condition'using safety grade equipment only."-

DSEROpenhssue-56). ' Minimum . cont' a inment : pressure' analysis for.. performance capability. studies:on the emergency core cooling system (ECCS)-(6.2.1.5).

~ Response:

. Westinghouse- does: not' plan. to~ take any action' unless the results of'the Staff's review are issued prior to the Final SER.

~DSER Open' Issue 57: Containment. pressure 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after accident (6.2.2).

Response

L-Westinghouse does not plan to take any action unless the results of the staff review are. issued prior to the Final SER..

DSER'Open Issue 58: Hydrogen purge and vent system.(6.2.5).

Resoonse:.

. Westinghouse does. not . plan to take any action unt4 additional guidance is received'from the' staff.

DSER Open Issue 59: Low-head pump deadheading (6.3.1).

Response

Use of the' low head pumps for tefety injection is not required for any size LOCA, since. the accumulators, high head safety injection (HHSI) pumps, and core reflood tanks are designed to perform this function. Nevertheless the low head pumps' are available for core cooling, either if containment spray is 1

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not' required 'from the start of the accident, or after the containment spray function has been terminsted.

Therefore, the operator may opt, at his discretion, for using the low head pumps for long -term core cooling. Normally, this would only be permitted l after stabh conditions have been achieved, i.e. several hours into the transient.- At that ' time, flow requirements are small (less than 100 lbs/sec)

.and ran be supplied by any one of the eight pumps; parallel operation of both pumps in'any one subsystem would, therefore, not be required.

j,- Even if both pumps in one subsystem were to operate in parallel, dead heading B

of the low head pump may not occur; this would primarily depend' on reactor coolant system pressure existing at the time. And even if dead heading were to occur, there would be no pump damage, because each low head pump is provided with a non-isolable miniflow path, including a miniflow heat exchanger, which is continuously supplied with component cooling water and which is adequately sized for this mode of operation.

Note that. a far more likely cause of dead headed operation would be the failure of the. closed motor operated valve in the containment spray line I (9011A) to open. Again, no damage would occur in this case because an i adequately sized miniflow is always available.

DSER Open Issue 60: Loss of-coolant accident (LOCA) analysis assumptions for ECCS(6.3.5).

Response

The ECCS flow assumptions used in the large LOCA analyses are as follows:

o The volume of one accumulator spills directly to the containment through the postulated break in one of the RCS cold legs. The three other accumulators deliver to the intact reactor coolant loop cold legs.

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o. Twofof the four high head safety injection (HHSI) pumps do not operate

* as a result of a' failure of one of the two emergency diesel generators to start. The . tao other HHSI pumps deliver to the direct vessel injection nozzles.

o All-.four of the .ccre reflood tanks (CRT's)- deliver. to the direct vessel injection nozzles.

The ~a bove ir in accordance- with conventional worst single failure' i assumptions. . In our cpinion, none of the content of Subsection 6.3.3 and/or Subsection 15.6.4 of RESAR-SP/90 PDA Module 1, " Primary Side Safeguards System," contradicts these analysis assumptions.

The statement in Subsection 6.3.2 of RESAR-SP/90 PDA Module 1 that any combination. of' five out eight HHSI pumps and CRT's is sufficient to meet the large LOCA functional requirements is related to a desire on the part of.

.Wr3tinghouse to provide additicnal margin in the design, that is to provide the integrated safeguards system with the capability to sustain - an additional single -failure. without a significant degradation in safety performance. In evaluating safety performance with an additional single failure, best-estimate methodology would be'used.

Note that this not only applies to the HHSI pumps /CRT's combination, but also

'to'the accumulators. These have been sized such that there is no. significant degradation in safety performance with only two-out-of-four accumulators delivering (i.e. one accumulator fails in addition to one accumulator L delivering) using best-estimate analysis methodology.

1 However, as stated before, in case of performing accident analyses, conventional single failure assumptions were used.

L l DSER Open Issue 61: ECCS flow to reactor vessel during LOCA/ compliance

'with Title 10 -to the Code of Federal Regulations, Part 50, Section 50.46 (10CFR50.46) and General Design Criterion (GDC) 35 (6.3.5).

- WAPWR-SER 15 MAY 1989 B197e:1d

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Response: e l

In general, Westinghouse does not establish. minimum flows for LOCA's for a ^

1arge number of break sizes; this would be impractical because it would .

requive an excessive amount of computer analysis.

instead, the following procedure is normally used in the case of a new design such as the SP/90:

(1) single.pointhighheadandlowheadflow requirements are established based 'on what have historically been the most limiting small and large>

LOCA's.

(2) safety injection system pumps and tanks are sized using the above flow requirements.

(3)LOCA sensitivity. analyses for a wide spectrum of breaks are performed to demonstrate that applicable criteria are met.

(4)' if applicable criteria are not met, sizes of safety injection pumps and tanks are adjusted and LOCA analysis are repeated as necessary.

In case of' the RESAR-SP/90 plant, the most limiting small break was judged to be the rupture of one of the direct vessel injection lines because it would '

leave only a ~s ingle high head safety injection (HHSI) pump available for delivery to the reactor coolant system. In addition, Westinghouse established an internal criteria that there should be no core uncovery for this particular small break. This conservative combination of sizing criteria led to a flow requirement of 65 lbm/see at 1000 psia, which in turn resulted in the pump described in RESAR-SP/90 PDA Module 1, " Primary Side Safeguards System,"

Table 6.3-2 (Sheet 1 of 6) and Figure 6.3-4; the capability of this pump to meet the above requirements is demonstrated in RESAR-SP/90 PDA Module 1 Figure- 15.6.4-36, which shows no core uncovery for a 4.313 inch diameter break i (this is the size of the direct vessel injection nozzle diameter), even

-assuming that the accumulators do not deliver.

WAPWR-SER 16 MAY 1989 B197e:1d

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.. L.c The most limiting large break was judged to be the cold leg double ended LOCA.. The flow requirement for thi's event was established as being able.to support a one (1) inch per second flooding rate plus margin to account for uncertainties and entrainment. In case of the -RESAR-SP/90, this led to a value of 643 lbm/see at 60 psia. This total flow had to be provided by the combination of two HHSI pumps and four core reflood tanks (CRT), except that an additional component failure was assumed (see the Westinghouse response to Issue 60 for additional background on this additional failure assumption).

The above flow requirement led directly to the CRT sizing shown in Table 6.3-2 (Sheet 3 of 6) of RESAR-SP/90 PDA Module'1.

m The . accumulators were scaled up from existing plants, except that also in this case additional conservatism was introduced by requiring that with only two out of four accumulators delivering there would be no significant degradation in safety performance (again refer to the Westinghouse response to Issue 60).

This'resulted in the' accumulators also described -in Table 6.3-2 (Sheet 3 of 6) of RESAR-SP/90 PDA Nodule 1.

Using the equipment sizes thus determined, the LOCA analyses for various break sizes as reported in Subsection 15.6 of RESAR-SP/90 PDA Module 1 were performed. . In all cases, there are large margins to the acceptance criteria described in 10CFR50.46, indicating that the safety injection portion of.the integrated safeguards system (ISS) delivers significantly more flow than the minimum required for each break size; it can, therefore, be concluded that the requirements of GDC 35 and 10CFR50.46 are fully met.

DSER Open Issue 62: Testing of fast transfer scheme (8.2.2).

Response

The FSAR revision is given below.

l The last paragraph of Criterion 18 (top of page 3.1-17, RESAR-SP/90 PDA Module 7, " Structural / Equipment Design") has been revised to read as follows:

1 WAPWR-SER 17 MAY 1989 B197e:1d w

y " Provisions for the. testing of Class IE AC electric power' systems, Class 1E DC. power' systems, 'and the standby power supplies.(diesel generators) l are described in Chapter 8 of RESAR-SP/90 PDA Module 9, " Instrumentation &

Controls and Electric Power."

, Provisions are provided for periodically testing-transfer of power among the-nuclear power unit, the offsite power system, and the onsite power system. This includes provision for . testing the fast transfer' feature described in Subsection 8.3.1.1.1 of RESAR-SP/90 PDA Module 9."

Inspection and testing of the offsite power systems are not included in the NPB  !

DSER Open Issue 63: Containment Electrical Penetrations'(8.4.1).

Response

Both RESAR-SP/90' (page. 6.5-30 of PDA Module 2,'" Regulatory Conformance") and the NRC Draft SER refer to outdated revisions of Regulatory Guide 1.63 and

'IEEE-317. The current revision of Regulatory Guide 1.63 is Revision 3 (February 1987). This revision accepts the requirements of IEEE 317-83 and

'IEEE' 741-86. The wording of the revision below is based on the current requirements. Details of penetration backup protection is in part dependent j on.' specific characteristics and details of vendor equipment installed in the L

plant. The revision is as follows:

The following has been inserted after the first paragraph of Subsection 8.3.1.4.1.3'(page 8.3-26) of RESAR-SP/90 PDA Module 9, " Instrumentation &

Controls and Electric Power":

"Both safety related and nonsafety-related electrical penetrations are protected against short-circuit damage. Frotective devices are selected and coordinated to ensure that rated short-circuit current and rated short-circuit thermal capacity is not exceeded. When a penetration WAPWR-SER 18 MAY 1989 B197e:Id

f . <. ' ' , - 'fg-p

. assembly. cannotL indefinitely withstand the maximum current available due L to~a".faultiinside containment backup. protection is provided. Backup

-protection consists of. dual. primary protection operating' separate interrupting devices, or primary cnd backup protection operating separate

~

interrupting. . devices. .The details of the containment electrical r penetration protectica (RG 1.63) are described in the following.

1. The only medium voltage power circuits passing through the eltetrical

^

penetrations are the reactor coolant pump-motor power feeders. Backup protection will be provided. by one of. two methods. One~ method provides backup protection through coordination, of pump feeder and bus supply breakers. A plant specific coordination study is required to verify =that this method is adequate. A second method provides backup protection via a second breaker in series. If this second method is utilized, the trip circuits will be supplied from separate battery systems.

2. For'480 volt- loads fed from load centers, primary protection is provided' by: the feeder breaker for the individual load. Backup

. protection is'provided by series fuses or by coordinated. protection with the load center bus supply breaker. Where a backup circuit breaker is used, the trip circuits will be supplied from separate battery systems.-  !

'3. For: 480 volt loads fed from motor control centers, primary protection is provided by the individual circuit breaker. Coordinated protection l with the motor control center supply breaker is the preferred method of backup protection. Protection for each circuit will be reviewed.

Where coordinated protection cannot be achieved, backup protection will be provided by an independent current limiting device installed with the supply breaker or by a separate backup breaker connected in series.

l l

l WAPWR-SER 19 MAY 1989 I B197e:1d l

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4. 125V::DC control: circuits: are protected by fuses and the system is y ungrounded. . Backup protection ~ is :provided by distribution supply

• . breakers. '

5. :120V AC' control circuits are low energy circuits which are protected by one fuse. -The available short circuit ~ current -for faults in the

containment ' isL generally-sufficiently low that backup protection is not required. Control circuits will be', analyzed and backup devices will be provided where required.

'6. Instrumentation ' circuits. are low energy circuits. The available short

[ circuit current- for faults in- the containment is generally sufficiently.- low that backup protection is. not- required.

Instrumentation' circuits will be analyzed and backup devices will .be provided where. required.

During" the' detailed design - phase, the exact method of overcurrent~

L

protection for each containment penetration conductor will be defined.

Containment : penetration overcurrent protective devices including backup.

devices which are required to.be operable will be identified in the FDA-submittal .."

DSER Open Issue 64: Power lockout to MOVs (8.4.3).

Response

Power Lockout to Motor-Operated Valves The following motor operated valves in the SP/90 design require power lockout to meet the single failure criteria.

< o Accumulator-Isolation Valves 8949 A, B, C, D o Core Reflood Tank Isolation Valves 9097 A, B, C, D

~WAPWR-SER 20 MAY 1989 5197e:1d '

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The -above information will be incorporated into the Technical Specifications for RESAR-SP/90 FDA submittal.

'DSER Open Issue 65: Reliability.to load sequencer with offsite power (8.4.7).

Response

The terminology " load sequence" in Subsections 8.3.1.1.2 and 8.3.1.1.3 of RESAR-SP/90 PDA Module- 9, " Instrumentation & Controls and Electric Power,"

refers to the load sequencing function and not to a hardware device.

The load sequencing function for the SP/90 is performed by the integrated protection system (IPS), in particular by the redundant trains of the engineered safety features actuation (ESFAC) subsystem. In case that safety limits for plant parameters are being approached, the ESFAC subsystem will generate system level ' actuation signals (see Subsection 7.1.1.1.2 of RESAR-SP/90 PDA Module 9, " Instrumentation & Controls and Electric Power" for a listing of these signals) which will be transmitted to the. integrated logic cabinets '(ILC's) which will in turn actuate the required protective components. These ESFAC signals will include the necessary sequencing information where applicable.

The ESFAC cabinets will be provided with the status of the Class IE electrical buses, and can therefore include sequencing information depending on whether or not offsite power is available.

This approach has several advantages:

o No additional hardware is required in the ESFAC cabinets to accomplish

.the sequencing function. A separate sequencer would be another component that could fail to perform as intended and would, therefore, reduce overall system reliability.

WAPWR-SEE 21 MAY 1989 B197e:1d I

.o The sequencing function shares in the increased redundancy provided by the ESFAC subsystem, i.e., the internal redundancy feature within each redundant train.

o The: sequencing function also shares in the IPS testing features, both in terms of automatic on-line testing and periodic system tests.

The reliability of the sequencing function will be demonstrated as part of the integrated protection system verification and validation efforts (FDA Open Issue #28).

DSER Open Issue 66: Station blackout (Unresolved Safety Issue [USI] A-44)

(8.4.8). Initially identified as Open Issue 5 of the staff's June 1988 Draft Safety Evaluation Report (DSER).

Response

This open issue has been addressed in Section 4.0 of RESAR-SP/90 PDA Module 2,  !

" Regulatory Conformance."

DSER Open Issue 67: Spatial separation of safety related systems (9.5.1.2.1).

Response

Subsection 9.5.1.1.d of RESAR-SP/90 PDA' Module 13, " Auxiliary Systems" has been modified as tollows to state the SP/90 separation philosophy more clearly and to make it consistent with NRC letter SECY-89-013 " Design Requirements Related to the Evolutionary Advanced Light Water Reactors (ALWRS)," dated January 19, 1989:

l

d. Within the Nuclear Power Block, redundant divisions of safety related equipment outside containment are located in redundant safety areas

which are separated from each other and from other areas in the plant by fire barriers with a minimum fire resistance rating of three WAPWR-SER 22 MAY 1989 B197e

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s , hoursi This~ degree of separation is'in many cases not. required, since-

.there are non-safety related components and systems in other fire

- .. am areas. that could be' relied upon to achieve safe shutdown following a fire- affecting = safety related. equipment; however, such separation

. addresses other, external events (e.g. flooding and sabotage) and greatly simplifies the design and analysis of the safety-related systems and is, therefore, cost effective. Each redundant safety area is further subdivided by internal fire barriers in order -to separate components which present 'a' fire hazard to other components or cable concentrations within' the same area. For example, each diesel generator is separated from .the. remainder of its associated safety area..

Exceptions to the use'of three hour fire barriers outside containment are ~only .made in'those cases where physical separation conflicts with other requirements or where the equipment is not clearly division 1 oriented. These exceptions are described in Subsection 9.5.1.3(a).

Outside' . containment, safe shutdown equipment is with very few

' exceptions division oriented, '.e. it can be clearly associated with' either Division A or Division B. Lis is not the case inside containment where only a few components fall into. this category. In addition, separation by three hour fire barriers inside containment is not practical because all compartments need to be in communication with each other in order to relieve pressure following a high energy line break. For this reason, the containment is considered to be a single fire area. Within this single fire area, separation of redundant shutdown equipment, including associated cables, will be designed to ensure, to the extent practical, that one shutdown division will remain free of fire damage.

1 WAPWR-SER 23 MAY 1989 B197e:1d

6 DSER Open Issue 68: Use of probabilistic risk assessment for. exemptions (9.5.1.2.1).

Response

PRA considerations are'an' implicit p' art of fire protection requirements, e.g.

the' assumption that ;a fire does not occur simultaneously with another unrelated event, or the fact that a single' failure need 'not be considered

, concurrent with a fire.

However, there is no intent'on the part of Westinghouse'to request exemption from fire protection requirements based on PRA, even if a detailed PRA for external events were to show that the risk due to. fire is small in comparison to the risk from other events.

DSER Open Issue 69: Deviations to 3-hour fire-barrier separation criteria

( 9. 5.1. 2.1 ) .

' Response:

Subsection '9.5.1.3.a of RESAR-SP/90 PDA Module 13, " Auxiliary System," has been modified to provide clarification with regard to the exception to the l general SP/90 separation criteria. With regard to the staff's questions on f

valve operation in the main steam tunnel (MST), the following response is provided.-

Question: Could fire render these valves inoperable, either open or closed?

Answer: Active valves' located in the MST are normally deenergized and i 1

require a signal to change position. If the cable carrying this signal. is damaged as a result of a fire, there is a high probability that the valve may be inoperable and will remain in its original position.

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l WAPWR-SER 24 MAY 1989 E197e:Id

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Question: .Could fire 1cause spurious action of these valves?

.3-~

jAnswer: -Since these; valves require an active . signal to; changeJ position,

~ '

spurious ' action of.these' valves is not considered credible.

N

]

Question:l If the answer to either or both of,these questions is "yes," could-4

.the. conditions cause or' lead to unacceptable -consequences

'vis-a-vis safe _ shutdown'of the plant.

' Answer:. As. explained in .more detail in the modified Subsection 9.5.1.3.a,.

inability to operate valves in the MST does not inhibit the. safe O- shutdown. capability of the plant.

1 l

. The revised Subsection 9.5.1.3 in RESAR-SP/90 PDA Module 13 follows:

4-l9.5.1 3 Protection of Safe Shutdown Related Equipment

. 1 a '. Separation'of Safe Shutdown Equipment

~

Safety-related, redundant components which.may be required to function following a fire in order to achieve safe shutdown will be protected as described in, Subsection 9.5.1.1(d).

Outside containment, all safety-related, redundant components are in principle located in two areas which are separated from each other and from other areas in the plant by fire barriers with a minimum fire resistance rating of three hours.

Safety area A contains Division A equipment. (mechanical, electrical and HVAC) as well as Channels I and III of the integrated protection system (IPS), including associated power supplies. Within safety

, area A, Division A and Channel I are combined into one separation gro9p, while Channel III~ constitutes another separation group. Separ-ation between these two groups in safety area A is in a:cordance with the pro.isions of IEEE Standard 384-1981. Similarly, safety area B contains Division B equipment as well as IPS Channels 11 and IV, with internal separation handled in a mannar identical to safety area A.

> WAPWR-SER 25 MAY 1989 B197e:1d c_ __ _

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. Indrdertominimizethepotentialforsmoke.and/or' fire, propagation,

there 'are: no?HVAC. systems.. serving both safety areas; .thus,' duct

,, - penetrations; between .the. two < safety ~ areas, which would require

, automatic = fire dampers, have' been eliminated. Similarly, fluid systems have been designed without cross connections between-reduddant-J safetyL divisions and therefore, there. are no piping connections g - between.the-two safety areas. . Cabling . penetrations through the fire barrier -between' the two safety areas are limited to the multiplexed,,

fiber-opt'ic; data links between ~the integrated . protection and A ' engineered safety features' actuation, cabinets'of the IPS.

Each redundant safety area is further subdivided by fire barriers; the.

. rating.of_these fire' barriers will be determined as the basis of a

~

s Fire Hazard Analysis to be performed during the detailed design phase.. .;

i

, All of. the equipment located in the redundant' safety areas can clearly 3 be: identified with redundant divisions of safety related systems and

- can, therefore, be clearly separated. .However, there is also safety related Lequipment located outside containment which cannot 'be separated by three hour: fire barriers, either because that would conflict with other requirements (i.e. the main control. room) o r. . ,

- because the equipment is not division oriented (i.e. the safety class portions of the main steam and feedwater lines located in the main i

steamtunnel).

The main control room (MCR) is an area where by.'its nature and intended use, multiple, redundant Class IE circuits and functions must exist within close proximity to one another. The approach used to deal with fires in'other parts of the plant, mainly fire barriers and/or separation. by distance, is not practical since to do so would inhibit the functionality of the MCR. Other methods must be adopted i which, as in 'other design areas, utilize defense-in-depth principles.

They also rely in part on the important feature that the MCR is continuously staffed. The following discussion addresses the RESAR

, SP/90 approach on three levels: prevention, detection, and mitigation.

WAPWR-SER 26 MAY 1989 B197e:Id-C-______._.______ _ _ _ . . _ . _ _ . _ . _

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,4 s , The: probability 1of.a fire is reduced through'a variety of methods:

.o -Low level voltage .on the switches (24V DC typically) reduces the opportunities for. fire to be initiated due to the' inherently- low

, , power levels available in the cabling.

o Multiplexing'significantly reduces the number of cables entering .

and ' leaving the .MCR, thus reducing the number: of sites for initiation, d ,

o 'Moreover, cables carrying multiplexed- signals use even lower b signa 1Lvoltages and carry miniscule power.

, 'o. Materials used in construction'of the panels' are inherently fire retardant. This includes the structures (steel), the surface coatings (fire retardant paint) and the. cabling .(fire retardant cabling).

These features ~ collectively' reduce the likelihood of-fire initiation and~ spreading beyond the local area should one start. In the event of a fire, rapid. detection is likely because of'the following:

o Plant operating personnel is continually inhabiting and observing the MCR.-

o Smoke detectors are installed, both inside the various panels and-in the general room area.

Mitigation'of the effects of.a fire is handled in several ways:

o Physical separation and/or fireproof barriers, in accordance with l

the requirements of Regulatory Guide 1.75, are incorporated in the panels 'with the objective of limiting the fire to one separation .

group.

WAPWR-SER 27 MAY 1989 B197e:1d 4

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l o' Portable. fire extinguishers are' included in the MCR.

o A firewater hose station is located directly outside the MCR.

o Breathing apparatus is provided in the MCR.

Because of the features described above, the probability of a major -

fire in the NCR is low. Nevertheless, the occurrence of such a fire

' J with . attendant MCR evacuation is postulated. For that purpose two emergency panel rooms are provided for in the SP/90 design. Each of these 'two emergency panel rooms is located in one of the redundant i safety related areas and provides the capability to bring the reactor to cold shutdown. For that purpose, the emergency panel rooms include

' indication of vital parameters and control of components required for cold shutdown.

The. safety class portions of the main steam and feedwater systems are not division oriented and are therefore not located in the redundant safety area; instead they are located in the main steam and feedwater tunnel (MST), which is situated between the two safety areas. The MST constitutes a single. fire area which is separated by three-hour fire barriers from the two safety areas; it is also provided with a separate and dedicated ventilation system. Within the MST, separation of redundant shutdown equipment, including associated cables, will be

~

designed to ensure, to the extent practical, that one shutdown division will remain free of fire damage even though safe shutdown equipment in the MST is backed up by equipment located in other fire areas.

The components located in the MST are shown on REE.'.R-SP/90 PDA Module t 6 and 8, " Secondary Side Safeguards System / Steam and Power Conversion System," Figure 10.3-1 (4 Sheets). The probability of fire initiation is low because of the absence of combustible materials and because of the small amount of cables in this area; furthermore, the majority of WAPWR-SER 28 MAY 1989 B197e:1d

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, the cables.normally do not carry voltage (valves in the MST need to' be energized to change. position).

Active . components in. the :MST .that could be used to effeet safe shutdown are:

1. -The startup feedwater control valves (1-FCV-1905 through 1-FCV-1908) which control the flow of startup feedwater to the steam generators.

2. The main steam power operated relief valves (1-PCV-1964- through <

1-PCV-1967) which are used to dump steam to atmosphere in case the '!

I

i main condenser is unavailable.

In case a postulated fire'in the MST were to disable these valves, the

-following . equipment located in other fire areas would be available to perform a similar function and to achieve safe shutdown:

1. The emergency feedwater system described in RESAR-SP/90 PDA Module .

6 and 8, " Secondary Side Safeguards System / Steam ~ and Power-Conversion System," Subsection 10,1,1.11.

2. As stated above, steam relief would normally be to 'the ma'n  ;

condenser. In the extremely unlikely event of a fire in the MST, j disabling the main' steam power operated relief valves coincident with the main condenser being unavailable, the steam generator overfill protection valves (1-9783A through 1-9783D and 1-9784A through 1-9784D) inside containment would be available to effect steam relief from the steam generators.  :

I In summary, equ'pment located in the MST that could be used as one metu for echieving. safe shutdown is backed up by equipment located outside the MST such that a fire in the MST will not prevent the operators from bringing the plant to a cold shutdown condition.

WAPWR-SER 29 MAY 1989 B197e:1d

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DSER Open Issue 70: Passive fire' protection features (9.5.1.2.2).

Response

-It is the intent of Westinghouse to select, whenever possible, passive fire protection components of proven design, which have previously been tested and

-are listed by nationally recognized testing laboratories.. In the unlikely Levent that . existing. designs cannot meet the functional requirements imposed by

. Wwestinghouse,.new designs.will be developed. Testing of such new designs; will be ' performed ;by, one of the nationally recognized laboratories in accordance with accepted standards in use at the time.

E DSER Open Issue 71: Water supply valve supervision (9.5.1.3.3).

Response

Westinghouse confirms that control'and sectionalizing valves in the fire water system will be,-electrically supervised and will be indicated in the m'ain control room. 1

-DSER Open Issue 72: Effect of. Earth. quakes on water supply system (9.5.1.3.3).

Response

Westinghouse confirms that the piping systems serving the standpipes and hose connections for manual fire-fighting, in areas containing equipment required for safe plant shutdown in the event of a safe shutdown earthquake, will be analyzed for SSE loading, and will be provided with supports to ensure system pressure integrity during SSE.

i WAPWR-SER 30 MAY 1989 B197eild

DSER Open Issue 73: Floor penetrations (9.5.1.3.3).

Response

The RESAR-SP/90 will not include floor penetrations that are susceptible to the potential of channeling water from fire extinguishing operations in one redundant fire area to an adjacent fire area. Floor penetrations will only be used for interconnections within one train of safe shutdown equipment; this is possible because of the highly separated layout of safe chutdown equipment in the RESAR-SP/90 plant.

DSER Ooen Issue 74: Reactor coolant pump oil collection system (9.5.1.4.4).

Response

Westinghouse confirms that the collection container is designed to hold the

)'

entire lube oil inventory of the reactor coolant pump motor. The largest potential oil leak referred to in Subsection 9.5.1.4.5 of RESAR-SP/90 PDA Module 13, " Auxiliary Systems," is used for drain line sizing purposes only.

DSER Open Issue y : Smoke control (9.5.1.4.5).

Response

The general arrangement of the RESAR-SP/90 safe-shutdown trains features a high degree of separation with no piping and very few cabling interconnee-tions. With such a physical arrangement, the ventilation system can become the most likely pathway for fire propagation and smoke dispersal; with this in

, mind, a decision was made for the RESAR-SP/90 to employ separate, dedicated ventilation systems for each of the two areas containing redundant trains of safe shutdown equipment, thereby greatly increasing the effectiveness of the fire barrier between these two areas. This arrangement of the ventilation systems serving the areas containing safe shutdown equipment will facilitate 1

l l

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l WAPWR-SER 31 MAY 1989 l

5197e:1d L __ _

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4 (the venting. of smoke originating in one area containing safe-shutdown

<: equipment,{and, will effectively ~ preclude spreading of this smoke to the i < redundant' area containing safe-shutdown equipment.

DSER Open Issue 76: Lighting systems (9.5.3).

' Response:

. Westinghouse has been notified 1 by staff that th'is is.no longer considered an open issue and will be closed out in the Final SER.

-DSER' Open Issue 77:- Emergency feedwater system (EFWS) actuation

. instrumentation and control (10.4.9).

R;esponse:

This open issue involves a continuing review on the part of the staff,'and Westinghouse will take no action unless the results of this review are issued prior to the Final SER. Westinghouse's review of'EFWS actuation indicates Lthat the requirements of GDC 19, the guidelines. of BTP RSB 5-1 andthe recommendations of NUREG-0611 are met.

DSER Open Issue 78: Compliance of EFWS with single-failure criteria (10.4.9).

Response

This .open issue -involves- a continuing review on the part of the staff, and Westinghouse will take no action unless the results of this review are issued prior to the Final SER. Westinghouse's review of the EFWS indicates that the requirements of GDC 34 and 44, and the recommendations of NUREG-0611 with regard to single failure are met.

e

'~ WAPWR-SER S2 MAY 1989 B197e:1d l

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DSER Open Issue 79: Source term calculations (11.1, 11.2, 11.3).

Response

Westinghouse has recalculated the RESAR-SP/90 " normal" radiation sources based on the methodology presented in ANSI /ANS 18.1-1984 (American National Standard Radiation Source Term for Normal Operation of Light Water Reactors). Revised Tables 11.1, 11.2 and 11.4 of RESAR-SP/90 PDA Module 12, " Waste Management,"

are supplied (as Attachment A) as requested for staff's use in their independent evaluation of the Westinghouse source term model. These tables will be updated by Amendment 3 to RESAR-SP/90 PDA Module 12. j DSER Option Issue 80: Compliance of the gaseous waste processing system with 1 10CFR50.34a, G0Cs 60 and 61, and RG1.140 (11.3).  !

l

Response

4 Westinghouse has modified Subsection 11.3 (and 11.2) of RESAR-SP/90 PDA Module 12, " Waste Management" to show that the gaseous waste processing system (and liquid waste processing system) meets the requirements of General Design Criteria 60 and 61, and the requirements of 10CFR50.34a.

However, the guidelines of Reguistory Guide 1.140 are for normal ventilation exhaust systems and are not applicable to the SP/90 gaseous waste processing system. As noted in the SRP, reference to R.G. 1.140 is made as an acceptance criteria, "as it relates to the design, testing and maintenance of normal ventilation exhaust systems."

DSER Open Issue 81: Sample capabilities for process systems (11.5). L

Response

The condenser air ejector stream is monitored prior to release and employs an automatic control function as stated in Subsection 11.5.5, Item G of RESAR-SP/90 PDA Module 12, " Waste Management," and as required by SRP 11.5 WAPWR-SER 33 MAY 1989 B197e:1d

r. .w

.. o. ,

, R l..

p

? . Table 1. However, a process ' system (i.e, the' main condenser air removal b . system -

NCARS) is? not' considered as part of the Nuclear Power Block (NPB),

h, but is the responsibility of the plant' specific applicant. This is supported a- .by the'_ staff's assessment in Section 11 3 of 'the Draft Safety Evaluation Report (DSER) which states, "The plant specific applicant will be responsible  ;

for- the radwaste building ventilation system and the turbine building ventilation system. Air and other non-condensable gases from the condenser .

l-

\ will 'be removed by.the main condenser air removal system. The plant specific applicant will be responsible for its design."

Additionally, the other ventilation systems identified in the DSER as not ,

having . provisions for grab sampling are also considered to be the responsi-bility of the plant specific applicant.- k' westinghouse retognizes that in process grab' sample provisions should be made for these systems according to Table 1 of SRP 11.5 and will add interface critoria to ensure the plant

~

-specific applicant will provide grab sampling capability as required.

l WAPWR-SER 34 MAY 1989 B197e:.1d

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.,. [ CATTACHMENT A (Sheet'2).

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j. TABLE 11.1-1 (SHEET 2 0F 2)~

l .

PARAMETERS USED TO DESCRIBE THE REACTOR SYSTEM-REALISTIC BASIS

[' ,

Isotope -Strippino Fraction

.. Kr-85m. 0.68 Kr-85 0.32-Kr-87 0.86-Kr-88 '0.75 Xe-131m 0.26

- Xe-133m 0.3b

^

Xe-133 0.27 Xe 135m :0.95' XL -135

~

0.48 ke-137 0.99 Xe-138 0.96 The nuclide stripping fractions are calculated using the following equation:

I " I ~ KQ:+ A (K .+ V) + P where:

Y = nuclide volume control tank stripping fraction.

- RT K = g.

m R = gas constant.(45.59 ag 9 ,, ),

T = nominal volume control-tank temperature (590'R). .

M = molecular weight of water (18.0 g/g mol).

H = Henry's Law constant equal to 3.3 x 10* atm/mol f raction for krypton 4

and 2.4 x 10 atm/mol f mt'lon for xenon, at 130*F 3

0 = letdown or purification flowrate (6.4 x 10 g/sec).

1 = nuclide decay constant (sec~I).

L = volume control' tank liquid mass-(6.7 x 106g,)

V = volume control tank vapor volume (1.0 x 10 7cm )3 P = volume control tank purge rate to the gaseous waste management system, at volume control tank conditions (1.3 x 10 2 g ,3/sec).

4 i

WAPWR-WM 11.1-6 AMENDMENT 3 l 3133e:1d MAY,1989

.mA_.m_m.u_. m___m.

m T;n g .% ,

Ex *1 NN0*: , , , TATTACHMENTA'(Sh5et3)

'j r .

t .

4 1

. . . I h . TABLE 11.1-2 (SHEET,1 0F 3) i SPECIFIC ACTIVITIES'IN PRINCIPAL FLUID STREAMS-REALISTIC BASIS 4

Normal Operation Source Terms (Based on ANSI /ANS-18.1 - 1984) -

8

[ . Volume Control Tank Purge of 1.30 x 102 cm3f3 t'

(8ASED ON ANSI /ANS-18.1-1984)- q GROUP I - NOBLE GASES

-Reactor S/G S/G LC oolant Liquid Steam.

Activity Activity Activity i L Nuclide- (vCi/a) (vCi/a) (vCi/a)  !

Kr-85m- 1.2E-01 Nil 2.3E-08 Kr-85 1.6E-02 Nil 2.9E-09 i Kr-87 1.4E-01 Nil 2.4E-08 Kr-88 2.3E-01 Nil 4.3E-08 4

Xe-131m 1.~1 E-01 Nil 1.9E-08 Xe-133m 2.7E-02 Nil 5.1E-09 1 Xe-133 '6.3E-01 Nil 1.1E-07 Xe-135m 1.3E-01 Nil 2.4E-08 Xe-135 5.9E-01 Nil 1.1E-07 Xe-137 3.4E-02 Nil 6.3E-09 Xe-138 1.2E-01 Nil 2.2E-08 GROUP II - HALOGESS Reactor S/G S/G Coolant Liquid Steam Activity Activity Activity

.Nuclide (vCi/a) (vCi/a) (vCi/a) __

I 6.5E-10 Br-84 1.6E-02 6.5E-08 I-131. 4.3E-02 2.4E-06 2.4E-08

. I-132 2.1E-01 3.0E-06 3.0E-08 I-133 1.4E-01 5.8E-06 5.8E-08

' I-134 3.4E-01 2.1E-06 2.1E-08 I-135 2.6E-01 7.2E-06 7.2E-08 WAPWR-WM 11.1-7 AMENDMENT 3

. 3133e:1d MAY, 1989

.. ow +.

. :e ,. .

ATTACHMENTAi(Sheet 4)-

=

TABLE'11.1-2 (SHEET 2 0F 3)

~ SPECIFIC ACTIVITIES IN PRINCIPAL FLUID STREAMS-REALISTIC 8 ASIS GROUP III RU8IDIUM AND CESIUM-Reactor - S/G S/G Coolant Liquid ' Steam Activity Activity Activity Nuclide fuci/a) (vCi/c) (uCi/c)

Rb-88 1.9E-01 4.5E-07 2.2E-09 Cs-134 6.4E-03 3.7E-07 1.9E-09 Cs-136. 7.8E-04 4.5E-08 2.2E ,

. Cs-137 8.4E-03 4.9E-07 2.5E-09 GROUP IV - NITROGEN-16 Reactor S/G S/G'

~

Coolant. Liquid Steam Activity. Activity Activity Nuclide (vCi/a) (vCi/c)- (vCi/a)

N-16 4.0E+01 8.2E-07 8.2E-081 GROUP V - TRITIUM Reactor S/G .S/G Coolant Liquid Steam Activity Activity Activity Nuclide (vCi/a) IgCi/a) (90i/0)

H-3 1.0E+00 1.0E-03 1.0E-03

,e l

WAPWR-WM 11.1-8 AMENDMENT 3

-'3133e:1d MAY,1989

_s _ _ __ . . _ _ _ _ _ _ _ _

%kgiWW - .

J o .; , '

,  : ATTACHMENT.~A(Sheet 5):

E f A . .

TABLE '11.1-2 (SHEET 3 0F 3)-

gy LSPECIFIC AC1IVITIES IN PRINCIPAL ~ FLUID STREAMS-REALISTIC BASIS GROUP VI - OTHER'IS0 TOPES ,

Reactor .S/G S/G

' Coolant. Liquid- Steam Activity. Activity Activity-fuci/a) (uci/a)

Nuclide (uti/c) 7 Na-24~ 4.5E 1.7E-06 8.7E-09 C r-51 , 2.8E-03 1.6E-07 8. 0E-10

'Mn-54 1.5E-03 8.3E-08 4.2E-10 Fe-55 1.1E-03 6.2E-08 3. 2E 'Fe-59' 2.7E-04 1.5E-08 7.7E-11

.00-58 4.2E 2.4E-07 1.2E-09 Co-60, '

4.8E-04 2.8E-08 1.4E-10 2n-65 4.6E-04 2.7E-08 1.3E-10 LSr-89 1.3E ~.2E-09 3.7E-11 Sr-90 1.1E-05 6. 2 E-10 3.2E-12.

Sr-91 9.2E-04 3.1E-08 1.6E .Y-91m' 4.6E '2.8E-09 1.4E-11 Y-91 :4.7E-06 2.7E-10 1.4E-12 Y-93 :4.0E-03. 1.3E-07 6.8E-10 Zr-95' 3.5E-04 2.0E-08 1.0E-10 Nb-95 2.5E-04 1.4E-08 7.2E-11'

. Mo-99 5.9E-03 3.1E-07 1.5E-09 1c-99m 4.6E-03 1.2E-07 6.0E-10 Ru-103 6.8E 3.9E-07 2.0E-09 Ru-106 8.2E-02 4.7E-06 2.3E-08 Ag-110m 1 1.2E 6.7E-08 3.4 E-10 Te-129m 1.7E-04 9.9E-09 4.9E-11 l e-129 - 2.4E-02' '

2.0E-07 ' 9. 9 E-10 Te-131m 1.4E-03 6.5E-08 3.3E-10 Te-131 7.8E-03 2.5E-08 1.3E-10 Te-132 1.6E-03 8.2E-08 4.1E-10

. Ba-137m 8.0E-03 4.7E-07 2.3E-09 Ba-140. 1.2E-02 6.6E-07 3.3E-09 La-140- 2.3E-02 1.1E 5.6E Ce-141 1.4E-04 7.7E-09 3.9E-11 Ce-143 2.6E-03 1.2E-07 6.2E-10

- Ce-144 3.6E-03 2.0E-07 1.0E-09

. W-187 2.4E-03 1.0E-07 5.2E-10 1 Np-239 2.0E-03 1.0E-07 5. 2 E-10

=

0 MAPWR-WM 11.1-9 AMUvDMENT 3 3133e:1d' MAY,1989

j~ y.y ...,x,

~-

1

.

• s j ,' g - l 89 JJ' ,

l , ' ATTACHMENT A.(Sheet-'6) y

.w TABLE 11.1-3 TRITIUM PRODUCTION (*)

Release Expected to Total Produced Reactor Coolant Tritium Source (Ci/ cycle) (Ci/ cycle) 1 Ternary fissions Initial cycle 23,520 2352 L

Equilibrium cycle- 23,520 2352 Coolant (soluble boron)

Initial cycle 420 420 Equilibrium cycle ,

530 530 Coolant (lithium,--deuterium,)-

}

1 Initial cycle 338 338 Equilibrium cycle 338 338 Total initial cycle 16,500 .3110 Total equilibrium cycle 11,400 3220 i

a. ThiToT1'owing. parameters were used:

Power level, 4200 MWt Release f raction f rom fuel,10 percent Lithium concentration (99.9 atom-percent 11thium-7), 2.2 ppm

~

. Initial cycle operating time,-13140 ef fective full-power h 1 Equilibrium cycle operating time,13140 ef fective full-power h U-I 1

6 WAPWR-WM 11.1-10 AMENDMEN1 3 3133e:1d. MAY,1989

$ylAg* , , f ' *o^ * ; ~ '

o ..

ATTACHMENT A-(Sheet'7)-

e i; '

s .-

TABLE 11.1 E.

,Di' ' VOLUME' CONTROL TANK k ' VAPOR SPACE ACTIVITIES c:

Activity Nuclide (uCi/a)

Kr-85m 8.9E-01 Kr-85 2 4E-01 Kr287 4.2E-01

.., Kr-88: 1.3E+00 Xe-131m 1;3E+00 Xe-133m 3.0E-01 Xe-133 7.3E+00

. Xe-135m 9.4E-02 Xe-135 4.9E+00 Xe-137 6.6E-03 Xe-138 8.0E-02 c

pc

.WAPWR-WM 11.1-11 AMENDMENT 3

' 3133e:ld : MAY, 1989

..