ML20128B050

Forwards Response to NRC 920805 Request for Addl Info to Continue Review of Application for Amends to Licenses NPF-9 & NPF-17,changing TS 3/4.7.5, Standby Nuclear Svc Water Pond, Transmitted Via Util

ML20128B050
Person / Time
Site:	McGuire, Mcguire
Issue date:	01/21/1993
From:	Mcmeekin T DUKE POWER CO.
To:	NRC
References
TAC-M79018, TAC-M79019, NUDOCS 9302020386
Download: ML20128B050 (19)

Text

_[.

Duke 1%ver Company :

T C Alan:N AlcCuire Nuclear Generation Department fire IWsident -

• 12700Hagers ferryRoad(AIG01A)

(704)STS-4800 lluntersville, NC280784983

- (741)875 48Q9 Fax DUKEPOWER January 21, 1993

Subject:

McGuire Nuclear Station

~

Docket Nos 50-369 and 50-370 Proposed Technical Specification Amendment Increase Allowable Temperature of the Standby l

Nuclear Service Water Pond (SNSWP)

Response to Request for Additional Information E

(TAC Nos. M79018 and M79019)

Dear sir:

By a letter dated August 5, 1992, the NRC staff requested additional information regarding their review of our proposed request to revise the Technical Specification 3/4.7.5 (Standby Nuclear Service Water Pond) that was submitted by a Duke Power letter dated October 23, 1990.

Accordingly, please find attached our response to your questions.

If you have any questions regarding our response or the amendment request, please contact Paul Guill at_(704).

875-4002.

Very truly yours,

/

f(HLMJ; T.

C. McMeekin

/

9302020386'9dO121 PDR ADOCK 05000369

/

i P

PDR 1 g\\

n-., --

4 U.

S.

Nuclear Regulatory Commission

+

January 21, 1993 page 2 xc:

Mr.

S. D. Ebneter Regional Administrator, Region II U.

S. Nuclear Regulatory Commission 101 Marietta Street, NW, Suite 2900 Atlanta, Georgia 30323 Mr.

P. K. Van Doorn Senior NRC Resident Inspector, McGuire McGuire Nuclear Station Mr.

T.

A.

Reed, Project Manager Office of Nuclear Reactor Regulation U.

S. Nuclear Regulatory Commission One White Flint North, Mail Stop 9H3 Washington, D.C.

20555

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9

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s V E I

ATTACHMENT

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t DUKE-POWER COMPANY' MCGUIRE N U C L E A R STATION 1

- 4 r

PROPOSED TECHNICAL SPECIFICATION AMENDMENT REQUEST

STANDBY.' NUCLEAR SERVICE WATER POND =

TEMPERATURE MONITORING REQUIREMENTS RESPONSE TO NRC. REQUEST FOR ADDITIONAL INFORMATION-4 m.

L 4

h 9

c 5

m I

s o'e'

e' NRC REQUEST 1:

FSAR-.page 6_162 indicates that the containment spray pump runout-

~

flow is 4800 gpm, while FSAR page 6-553 indicates that containment spray pump runout is 3400-gpm.- which.value is correct _and which-was used in the containment peak pressure analysis?

How does this value relate to Table 6.2.1-18?

DUKE'S RESPONSE 1:

The correct value for the containment spray pump runout flow is 4800 gpm.

The design flow is 3400 gpm, which is the value used in the containment peak pressure analysis.

Table 6-126 describes the steps and_ changes that occur in RWST flows and RWST inventory during switchover to the recirculation mode, and has no relation to the peak containment analysis.-.The pump flows for this table were provided by Westinghouse in a letter dated August 24, 1979 as the-maximum flows for RWST level calculations.

The flows-for RHR, SI, and CC syrtems are runout-flows for a two pump configurations.

The spray-pump flow is the design flow, since the spray pump would operate against its normal-system head even in case of a LOCA.

9

NRC REQUEST 2:

I FSAR page 6-553 indicates that the runout flow for the RHR-pump is-4500 gpm which is inconsistent with other data available.

What is the correct RHR pump Runout flow?

Provide a pump curve (including a system curve) to show the design point and actual operating point for the RHR pump.

DUKE'S RESPONSE 2:

The correct runout flow for the RHR pumps is 4500 gpm.

Enclosure 1 provides the pump curves for each RHR pump at McGuire.

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- - NRC REQUEST 3:

DPC - has indicated _that site _ functional tests were performed - to determine the RHR pump flow during_the recirculation phase ~of a postulated loss-of-coolant accident (LOCA),

What. were the objectives and results obtained from the site functional testing (i.e.,

flow rates, flow resistance, etc. ) ? With_ approximately 250_0 gpm diverted to. auxiliary spray, _what is the flow (measured or-calculated) remaining for ECCS during the recirculation portion of the LOCA?

DUKE'S RESPONSE 3:

In 1978 a site functional test was performed to verify the flow paths from the RHR pumps to the reactor vessel through the safety injection alignments; to verify that the flow rate and pump l head are above the minimum required; and to verify that the RHR pumps provide adequate NPSH during various safety alignments.

Enclosure 2 is a report that documents the results of the functional test.

Please note that the functional test was only concerned with determining flow, thus no other parameters were determined. By a May 10, 1979 letter, in response to our request, Westinghouse transmitted their comments on

'h functional test that was performed (Enclosure 3).

DPC currently takes credit for 2400 gpm auxiliary containment spray flow for the McGuire Nuclear Station.

This flowrate was developed by modeling sump recirculation with a single low head safety.

injection pump supplying flow to the auxiliary containment spray 4

header, 2 high head sefaty injection pumps, and 2 intermediate safety injection pumps. The 2 nign nead safety injection pumps and the 2 intermediate safety injection pumps are assumed to _ be operating at their respective maximum combined pump flowrates. The flow diverted to the ECCS is thus approximately 1800 gpm.

If only one train of the high and intermediate head safety injection systems is operational, the pumps are assumed to operate at runout flow _ conditions, and the flow available to the' ECCS _ is -

approximately 1300 gpm.

By_ assuming that the high and intermediate safety injection systems are operating at maximum flow conditions, the flow available for auxiliary containment spray is conservatively minimized.

The flowrates available for auxiliary containment and emergency core cooling are consistent with the 4500 gpm design flowrate of the low head safety. injection pumps.

i I

4

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0 0

0 ENCLOSURE 2 l

I i

w

~

TP/1/A/1200/03C Saf ety injection Pumps and Flou Adjustment-Functional Test'-

-(Data _and Results)

P,tre C _RHR Pump Over-Run

-Narrative Tha final section (12.5) of the test was to verify _the flow paths-from the ND

. pumps to the reactor vessel through the safety' injection alignments, verify tha flow rate and pump head are above the minimum required, and verify that ths ND pumps provide odequate NPSH during various safety alignments.

The pumps _

cut and exceed all test criteria.

. Thsre are no specific maximum flev values for any of the various alignments, over and beyond the-implied limit of the end of the pump curve in the FSAR

(Tigure 6.3.2-5) at 4500 gym.

In all but one case, the flow rates exceed 4500'gpm as shown in the following data, e

tc.e-1 A t,Nv)

T

'Y

< as, Lt.P-sa (uv) 4* "

T auR P.in (wo) aa 6

t wo so i

$Li up :

e 4 FU5T M t.

NI w3 cp no,g.

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"'M T

RHR P.16 tuo)

-+

T p r isi.

~Q

-pg Nf J 30 u t p.gg wr J34 "f.%3 5 l

A11gnewnts 1..* NDP-1A aligned to Cold Legs A and B NDP -1A aligned to Cold Legs A, B, C, and D (Cross-over open) 2.

3.

NDP-1B aligned to Cold Legs A and B

4..

NDP-1B aligned to Cold Lcge A, B, C and D (Cross-over open)

Data 1.

Flovrate 4507 gpm Pump Suction 35 psis Pump Discharge 182 psig Test Gauge at Discharge 180 psig 2.

Total flowrate 5196.4 spm Legs A and B 3044.0 g(t" Legs C and D (x over) 2152.4 ggn3 3.

Flowrate 4346.1 gpm Pump Suction 35 psig Pump Discharge 194 psig Test Gauge at Discharge 194 psig 4.

Total Flowrate 5418.8 gpm Legs C and D 2735.7 gpm Legs A and B (X over) 2683.1 gpm Alignments NDP-1A supplying suction of 2 CCP's, 2 NIP;s, and injecting into cold 1.

legs A and B.

NDP-1B supplying suction of 2 CCP's, 7 NIP's, and injecting Lato 4 cold 2.

legs.

Data 1.

NI Flow 630 gpm each NV Flow 795 gpm total ND Flow 3102 Total 5157 2.

NI Flow 625 gpm each NI Suction 15 psig NV Suction 20 psig NV Flow 807 ND Flow 3436.6 epe ITD Suction 14 psig Total 5493.6 gpm

I"

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4 e

4 4

ENCLOSURE 3 9

• f

" 00KE-4799

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f IT)Mgl4 I

%5dw=m omsim Weslingh00se Waler Reactor d

en u

[leClfl0 00fp0f 8110n Olvi$lonS PttitDJgh Pennsylvania 15730 MAY 10,1979 SD/SS-DAP-237 Mr. S. K. Blackley, Jr.

MPS #4100 Chief Engineer Mechanical and Nuclear Division DAP-280 pMCH-79M-6,1/11/79 E*"

Ref.t g

3 Charlotte, North Carolina 28242 McGUIRE NUCLEAR STATION UNITS NUMBER 1 AND 2 SIS /RHR Functional Tests Dear Mr. Blackleyt Duke Power, reference 1, transmitted the McGuire SIS and RHRS functional test results and requested Westinghouse's comments on the RHR pump runout flows. To summarize the tost results,the injection case with one RHR pump aligned to four cold leg injection lines resulted in RHR pump "A" delivering 5196 GPH or RHR pump "B" delivering 5419 GPH. The NPSH required at 5500 GPM is 24 feet, reference DAP "A" and "B" Ingersoll-Rand RHR pump Curve Nos. H-462, 4 while the calculated minimum NPSH available from the RWST is 46.

feet, therefore the RHR pump Injection runout flow is acceptable.

The cold leg recirculation case with RHR pump "A" aligned to two centrifugal charging pumps, two safety injection pumps and two cold leg injection lines resulted in a flow of 5157 GPH. The NPSH required at 5200 GPM Is 21 feet, while the calculated minimum NPSH available from the containment sump is 29 2 feet, therefore the RHR pun,p recircu-lation runout flow is acceptable.

It is noted that the minimum NPSH available values are calculated with conservative suction piping resistances and minimum RWST or sump water levels. The actual NPSH available values will add further margin to the Injection and recirculation cases discussed above.

Please contact the undersigned if additional clarification is raquired.

Very truly yours.

-NL/

R. 3. Howard, Manager DUKE Power l'rojects b

~

C. R. Sterrett/L. E. Conway/mme cc:

0. I.

Fuller 1L S. K. Blackley, Jr.

6L j

L. E. Veaver il L --

---

l-_--...,__--_,___l____,

NRC REQUEST 4:

Your March 3, 1992, response indicates that heat exchanger fouling has reduced the containment heat exchanger UA (from 2.94 to 1.47 E06 Btu /hr-F) and the component cooling water heat oxchanger UA (from 5.0 to 1.6 E06 Btu /hr-F).

This is a substantial change that should af fect the analysis results. Please provide the significant analysis inputs and the corresponding analysis results for the containment peak pressure anslysis that supports the subject TS submittal.

Identify the inputs that were changed as part of the containment peak pressure analysis and to the extent possible (recognizing that when several analysis parameters are changed it becomes difficult to quantitatively identify the impact each 1

individual chango has on the calculated containment peak pressure) discuss the impact that analysis input changes had on the analysis results.

t DUKE'S RESPONSE 4:

The significant analysis inputs for the peak containment pressuro analysis are provided within Enclosure 4.

The input assumptions for LOTIC 1 analysos has been changed for various reasons several times in the course of a few years.

It is difficult to provido an exact quantitative effect of each change on the peak containment pressure, since sensitivity studios were not performed, and in most cases several input changes were made at the same time.

The inputs listed above the dotted line were changed in a

conservativo direction, causing the peak containment pressure to increase, or were not changed at all.

While those below the dotted lino resulted in a

decrease in -peak containment

pressure, compensating for the loss of margin of the group above the dotted line.

Especially significant was the change in tho. mass /onorgy release model that Westinghouse incorporated in LOTIC-1 in January 1986.

As'a result of the change, the peak containment prosauro went to 12.7 psig, compared with 14.8 psig calculated in the initial FSAR.

It should be noted that parallel _with this change, soveral other variables were modified in the conservativo direction.

Also significant was the increase in auxiliary spray flow by almost 48%.

Its individual effect cannot be determined, since other inputs were changed at the same time.

In conclusion, it was the combined offect of the last four inputs in the above table that gained margin and allowed significant reductions in heat exchangor UA values,

.m

.~4,

e 9

ENCLOSURE 4

~

Current 15AR Initial FSAR (with license application) 6 Initial ice Weight in the ice Condenser 1.89 x 10 lbm 2.22 x f(f Ibm Standby Nuclear Service Water Pond 820 F 78 4 Temperatere Containment Spray llX lleat Transfer 1.47 x 106 BtuAir T 2.94 x 1(f BtuAir.4 Coefficient:

Residual llcal Removal llX llent Transfer 1.64 x 106 Blu/hr DF 1,64 x 1(f Dtuhir.0F Coefficient Component C< cling IlX llcal Transfer 1.60 x 106 Blu/ht T 5.00 x 1(f Bluhir T Coefficient Containment Structural lleat Sink

-7 % less than See initial FSAR Areas initial FS AR Nuclear Senice Water Flow to Containment 3800 gpm

$000 Fpm Spray lleat Exchanger Nuclear Service Water Flow to Component

$500 gpm 8000 gpm Cooling fleat Exchanger Spray Flow from Refueling Water 3400 gpm 3400 gpm Storage Tank Spray 190w from Sump 3432 gpm 3432 gpm RWST Water temperature 105'F 120+F

~ Auxiliary Spray Flow from Sump 2400 gpm 1623 gpm (ND Spray)

Active Sump Volume 90.00() fil 46.500 ft3 Westinghouse hiass/ Energy hiodel Ref. 36(3)

Ref. 250)

(WCAP 10325)

. (WCAP.8312.A)

Peak Containment Pressure Calculated With LOTIC 1 14.07 psig 14.8 psig (1) Dese are references as listed in the current FSAR F

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