ML19309H421

Forwards Responses to NRC Info Requests 212.102,212.107 & 212.72 Re Fsar.Plant Will Employ Leakage Return & Leakage Detection Sys Which Assist in Identifying & Controlling post-LOCA ECCS Leakage

ML19309H421
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	05/02/1980
From:	Novarro J LONG ISLAND LIGHTING CO.
To:	Harold Denton Office of Nuclear Reactor Regulation
References
SNRC-473, NUDOCS 8005130207
Download: ML19309H421 (12)

Text

o 8005130/07 f

LONG ISLAND LIGHTING COM PANY

._ FLC'O jfag,w w w SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY RO AD e WADING RIVER, N.Y.11792 May 2, 1980 SNRC-473 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322

Dear Mr. Denton:

Forwarded herein are fifteen (15) copies of LILCO's responses to information requests 212.102 and 212.107 contained in the following letters:

1. NRC letter of September 19, 1978, S. A. Varga to A. W. Wofford.

2. NRC letter of November 15, 1979, L. S. Rubenstein to A. W. Wofford.

Also enclosed is a revised response to question 212.72, previously submitted in FSAR Revision 11 dated June, 1978.

We trust that the enclosed responses will provide adequate resolution to your concerns. Should you request further-information please do not hesitate to contact this office.

Very tr ly yours,

) '

hm%)

I . P. No arro Project hanage,r Shoreham Nuclear Power Station LG/cc i Enclosures h00l cc: J. Higgins

/

e 4 493!

't LONG ISLAND LIGIITING COMPANY Si!OREllAM NUCLEAR POWER STATION - UNIT 1 RESPONSES TO NRC QUESTIONS MAY 2, 1980 I

SNPS-1 FSAR l

i RESPONSE TO SECOND ROUND REQUESTS (SETS 21 & 25)

Text, tables, and figures referenced in these responses, and attached, contain new or revised material. Text, tables, and figures referenced, but not attached, are existing material as presented in the FSAR.

1 I

1 Docket No. 50-322

.l 1

4 s- - - .-

~

TABLE OF CONTENTS REQUEST NO. PAGE TABLES FIGURES 212.72 212.72 212.102 212.102 212.107 212.107 - 212.107b 212.107 212.107-3 l

l l

1 1

4

SNPS-1 FSAR Request 212.72 (6.3) :

The ECCS should retain its capability to cool the core in the event of a passive failure during the long-term recirculation cooling phase following an accident. Shoreham should determine the effects on ECCS of passive failures such as pump seals, valve ,

seals, and measurement devices. This analysis should address the potential for ECCS flooding and ECCS inoperability that could result from a depletion of suppression pool water inventory. The analysis should include consideration of (1) the flow path s of the radioactive fluid through floor drains, sump pump discharge piping, and the auxiliary building; (2) the operation of the auxiliary systems that would receive this radioactive fluid; (3) the ability of the leakage detection system to detect the passive failure; and (4) the ability of the operator to isolate the ECCS passive failure, including the case of an ECCS suction valve seal failure.

Also, examine the auxiliary system piping in the location of ECCS equipment and address the potential of nonsafety-grade pipe to cause flooding.

Response

The response is incorporated in the response to Request 212.107.

Also, details of the moderate energy crack analysis and the effect of failtre in auxiliary or nonsafety piping in the ECCS areas is discussed and analyzed in Section 3C.5 of Appendix 3C.

212 StiPS-1 FSAR Request 212.102 (6.3) :

With regard to the means to cope with a limited leak during the long-term post-LOCA period, we find that the Snoreham design does not provide acceptable protection. This conclusion is based upon our general observation that insufficient design features exist to detect and isolate postulated leakage of a pump seal or valve packing subsequent to the accident. In particular, the leakage detection system would not readily identify the ECCS train which has the leak. Individual ECCS trains raay need to be isolated in a trial-and-error method to locate the leak. Such gross procedures are not sufficiently precise to locate leaks within a reasonable time period, nor is the unnecessary cycling of operating core cooling systems desired as a planned leak-identification procedure. The sensitivity or the proposed leakage detection alarm is relatively low. Approximately 9,000 gallons (1 1/2 hours for a 100 gpm leak) are required to actuate the control room alarm. Other aspects of the proposed leakage capability which are also unacceptable include: (1) the potential for a seal failure in inboard valves (closest to the primary containment) in the ECCS suction lines, test lines, and minimum flow lines wnich cannot be isolated, (2) expected through-valve leakages for an extended period or time; (3) tiow patns of radioactive fluid through floor drains, sump pump discharge piping, auxiliary buildings, and auxiliary systems; (4) lack of documentation to show qualitication of the leak detection system to IEEE-279 (except for single failure) ; (5) need for inordinate operator action;. (6) sensitivity of the changing suppression pool level to detect a small leak.

It is 'our position that LILCO must provide design modifications to increase their ability to detect and 1solate leaxage subsequent to a LOCA. An acceptable modification would be to install water-tight pump rooms. Other alternatives or additions to consider would be enclosing inboard isolation valves and/or installing suitable curbing to partition selected leak-monitored areas. Whichever approach is selected, sufficient design detail and capabilities must be submitted to allow an adequate final evaluation.

Response

The response is incorporated in the response to Request 212.107.

212-102

. SNPS-1 FSAR l

Request 212.107-l With respect to the means to cope with passive failures during the long term post-LOCA period (refer to NRC Staff Question Nos.

212.72 and 212.102), provide appropriate plan and elevation drawings of the annular area between the suppression pool and the reactor building where ECCS equipment is located. The information should include equipment general arrangement drawings in this annular region, as well as piping isometric drawings which clearly distinguish to which of the safety systems the piping is associated with. Also show the location and elevation of sumps for draining water which has leaked into the annular area. It is particularly important that you complete your response to our previous Question Numbers 212.72 and 212.102. We need complete responses to these questions before we can complete our review in this area.

Response

The Shoreham plant will employ a leakage return and leakage detection system which will assist in identifying and controlling post LOCA ECCS leakage.

Redundant safety related level detectors will be provided on el 8-0, which will alarm in the control room when the floor water level (in the detector area) exceeds approximately 1/2 inch (corresponding to approximately 2000 gallons) . These detectors will be powered from emergency supplies, and are also employed in the moderate energy crack analysis, as described in FSAR Appendix 3C, Section 3C.S. In addition, there is a level detector in the floor drain sump which will alarm in the control room at high sump level.

The leakage return system vill consist of a self priming J eakage return pump, as shown on Fig. 212.107-1, with a capacity of 180 gpm which includes a recirculation flow of 50 gpm. This pump will be~ manually started as required and will operate to return postulated ECCS leakage to the suppression pool. The pump will be powered from the emergency supply and will be seismically qualified. In addition, the normal reactor building sump pumps will be automatically tripped on a LOCA signal to ensure l

radioactive leakage is not removed from the reactor building.

[

The use of the leakage return system during post LOCA conditions will allow sufficient time for operator action to identify and isolate suspected leakage paths while continuing to maintain suppression pool water inventory and preventing excessive buildup of water on el 8-0 of the reactor. building.

The ECCS systems are capable of withstanding passive failures following a LOCA, such as pump seal, instrument line, or valve 4 packing failure. The maximum leakage due to a failure of this nature could be as high as 100 gpm, due to.the break of a core spray pump discharge pressure instrument line. The maximum 212-107

. SNPS-1 FSAR leakage from any one low pressure ECCS pump seal f ailure would be 50 gpm. Valve stem leakage would be significantly less than either of these.

Should post LOCA ECCS leakage of 100 gpm occur in the reactor building the water level at the 8 foot elevation would be detected and alarmed in the control room within about 20 minutes.

The operator would then take necessary action to isolate the source of flooding as follows:

1. The operator will receive two alarms, one at floor level

" Suppression Pool Pumpback Reactor Building Floor Drain Tank Level High" and a second one-half inch above el 8-0

" Reactor Building Flood Level High". After receiving these alarms, the operator would attempt to ascertain and isolate the source of leakage using available system indications (e .g . , a broken instrument line may be detected by a failed-low or failed-high pressure indication) . At the same time, the operator would monitor the el 8-0 water level to determine if continuous leakage were present. If the level continued to increase, ECCS leakage would initially be assumed and the leakage return pump started to ensure no reduction in suppression pool inventory. This action alleviates the time constraint on leak detection and allows the operator to initiate selective identification / isolation procedures. Activation of the leakage return pump will occur long before reaching the approximate 22 inch level outlined in the moderate energy crack analysis as beginning to impact on essential equipment (see Section 3C.S.4.5) .

2. If the source of leakage is not immediately found, j sequential ECCS loop isolation / suppression pool level I monitoring would be required. The operator would-shut -

or check shut RCIC suction valves from the condensate storage tank and from the suppression pool. The corresponding valves on HPCI will have automatically shut earlier in long term post LOCA conditions and would be checked to be in the shut position.

3. The operator would now secure one LPCI/RHR loop and one core spray loop and would shut the corresponding pump suction valves.

4. The above isolation procedures may require the operator to intermittently operate the leakage return pump to observe the effects upon water buildup. When the leaking ECCS loop has been determined, the operator would return the unaffected ECCS trains to service (a s required) and isolate the leakage return pump from the  ;

containment.

212-107a 1

4

, , SNPS-1 FSAR The typical sequence above would ensure the isolation of any pump seal, valve packing, or instrument line failure in the ECOS systems following a design bases accident, with no effect on suppression pool water inventory or excessive water buildup in the reactor building. Any ECCS system leak would be isolated in the above sequence, including packing f ailure on any ECCS pump suction valve. This packing can be isolated, since the valves are double-seat, wedge knife gates.

The general scenario discussed above assumed a worst case 100 gpm leak during a design basis accident. The came basic approach would apply following other than DBA LOCA's and to smaller leaks.

For smaller leaks, level instrument response would be slower in all cases, ECCS equipment would be unaffected and total suppression pool inventory would be maintained.

This discussion is not intended to represent a detailed leak isolation procedure, but to demonstrate system design capability to withstand a worst-case leak.

The leakage return pump and level detection equipment could also assist in controlling post LOCA passive failures in auxiliary or nonsafety piping (in the ECCS area) until such time as the leak could be isolated by methods similar to those described above.

The effect of failure in auxiliary or nonsafety piping in the ECCS areas have been discussed and analyzed in Section 3C.5 of Appendix 3C. During normal plant operation, the existing reactor building sump pumps will assist in controlling (and detecting) any leakage to the ECCS area, until the source can be ' identified and isolated. Refer to Section 3C.5 of Appendix 3C for limiting moderate energy failure analysis. All leakage to the el 8-0 area will be routed through floor drains to one of the three reactor building sumps.

The el 8-0 area is a common ECCS area capable of storing large volumes of postulated leakage before impacting ECCS equipment.

Ieakage from any ECCS train in the area will be effectively returned to the suppression pool via the leakage return pump.

For reference, the ECCS piping and equipment at el 8-0 is shown on Figs. 212.107-2 and 3.

l i

212-107b i

I l Et t - 16"- WR -22 2-151 -2 J -

lE21 l f Elf EXISTING CORE g f E 21- 10"-:

W R 151- 2 IE11-16"WR-309-15F2 SPR AY TEST LINE X-LOA LO'W WATER CC3 :  : CC2 _ g EL. 2 6*-O" Pi _

640Ch 5MOV IG11 l *l E21-3"-W R-639C IE2f W LAH 3/4"-VO S -60 C- 117-151-3 L.C.

3"-DRW 151-3 d VCW-15A-2 L.O. OOL

• LIS s (PCM)  ; ; VGW- SA-3 RO 3"- D R W 15f -2 643C 641 C [ 3/4"- DRW- 25-151-2 FE 647C [

WLTS . D RW- 27-151-3

- l/2"VOS-60C0 (T Y P.)

h I f' I

,wVOS- 6 0 C- 2 (TY P.)

642 ChW L E t- h"- D R W-43-151-3 LO.

642 k l X 3/4 -DRW- 26-151-2 s

() VGW-15 A-3 "F T 4'iDRW-138-151-4 N IG11

• P -270 C 647C \ *pg 3"D RW- 23-151-3 647C

] g LEAKAGE RETURN PUMP (SELF-PRIMING) tEL. 8'-O"

1. 1. # W 77 4" - D R W- 22 -151-3 FIG. 212.107-1 LEAKAGE RETURN SYSTEM m SHOREHAM NUCLEAR POWER STATION-UNIT 1

(

j

\ IG11-P-036E C F FINAL SAFETY ANALYSIS REPORT IG11-TK-056C(FLOOR DRAIN SUMP)

1 1  !

l \

l 1E41-14"-W-4-901A-21

l 1E41-4"-W 9-901A l

. *~.

,,, .e.,. 'e

'e*P,*

IE!

1E21-14"-WR-1-151-2 ,' . e. . ' RC

,' RC CORE SPRAY PUMP A ' .' , (NOT

~

I 1E21+P-013A ~

1E11-3" WR-241-301 * . '

1E11-16"-WR-231-301-2 ,

IE21 14" WR-1-151-2 1E11-6"-WR-368151-2  ; -R PUMP IE11-6"-WR-341151-2 s j 1E11*P-014 A , I.

IE11-20"-WR-221-151-2 ',

FLOOR DRAIN ,,

SUMP h, s#

.- -1E21-3"-WR-4-301-2 ,

1G11-TK-056(C)- ~ 4[--l - IE11*MOV 044 A ._

l -IE11*MOV 045A 1E11-20'tWR-221-151-2-7L. ' - 1E11*MOV 032 A 1E11-4"-WR-372-301 -

LEAKAGE RETURN " -

PUMP RHR PUMP 1G11*P-270C , .. 1E11*P-014(C) 1E11 16"-WR -233-301-2 - , ;, ?

1 Ell-3"-WR-243-301-2 '.'#. ~

IE11-20' LWR-213-151-2 1E11-20"-WR-223-151-2 - ,'['. '

' ,1E11*MOV 032C LEGEND

(

RESIDUAL HEAT REMOVAL SYSTEM (1E11)

HIGH PRESSURE COOLANT INJECTION SYSTEM (1E41) g CORE SPRAY SYSTEM (1E21) 1

HCPIPUMPC. BOOSTER 1E 41*P-C16 1E41-16"-W 1-151-2 1E4I 4"-W-9-151-2 l

g _y -, , .

2-1E41-MOU O36/ _

r i)-6"-W-2-131-2

PUMP 1E51+P-015 1E51*MOV031 ..

1E21-14"-WR-21-151-2 JC PUMP SUCTION LINE PART OF ECCS PIPING) ,,s ,.

._e $ ' ,' . CORE SPRAY PUMP

, , , ; . e, 1E21*P-013B 1E21-14"-WR-21-151-2 ' ' ,, IE21-3"-WR-24-301-2

, IE11-4"-WR-336-301-2 1E21-12"-WR-22-301-2 J RHR-PUMP N - 1E11+P-014 B J

1 Ell-3"-WR-242-301-2 %

["E -- 1 Ell-16"-WR-329-301-2 t . _.

1E11-6"-WR-370-151 ~_d

--1E11 MOV 044 B 1E11-6" WR-342-151 -1 Ell-4"-WR-250-301-2 p '.- '

,,- 1E11*MOV 045B s '-

~1E11-16"-WR-234-301-2

.l.-'

• • IE11+MOV 032B'

~

1E11-20"-WR -232-151-2 1E11-3"-WR-244-301-2 1E11-20"-WR-212-151-2

..~

RHR PUMP

!E11*P-014(D) 1E11-20"-WR-214-151-2 . , ,, .

~

, , . 1E11-20"-WR-224-151-2

., . ' 1E11*MOV 032D

'(: ,' ' .-

'.- t, FIG. 212.107-2 ECCS COMPOSITE PIPING PL AN EL.8'-O" TO 20'-O" SHOREHAM NUCLEAR POWER STATION-UNIT I FINAL SAFETY ANALYSIS REPORT

1E 51-IE41-14"-W-3-9018-2

} \1E41-14"-W-44 5

, RCIC PUMP SUPPRESSIO

. POOL SUCTION LINE

• 9 (NOT PART OF ECCS PIP 1:? 'c'-

4--

g. 4 ,:* '

X-21A i~ O - 1E21-14'!WR-1-151-2 1E11-16'!WR-245-301-2 1E21-lO'!WR-7-151-2 RHR HEAT EXCHANGER fr (T  % 1E21-12"WR 2-301-2 1E11*E-034 A 1E1H6"-WR-253-301 +

gr q

f'-

)( -

7 IE11-16'tWR-3091512 .. e X-10A h1E11-18WR-247301-2

- X-9A 4V

~-A

.D 1E11-16"-WR 233-301 '

d [h1E11-20"WR-205-151-2 1E11* MOV O31A h .c.

i l IEll-20"-WR-211-151-2 1E11*MOV 035A -

\' ' -lE11-20"WR-204-151-2 IEll-20"-WR-221-151-2 '

S 1E11-20"WR-207-151-2

- X-9C k.. , .;3,

- 1E11*MOV 031C Ty.'. i.' i \

? 1E11-24"-WR-255-301-2 j IE11-20"WR-223-151 ' l

~

l v:.?-l4.,

' 5 l, 1

l l lEll-20#

NORMA MODEd (NOT U LEGEND ALOCq h RESIDUAL HEAT REMOVAL SYSTEM (IEll) l l l

'hfk;f p HIGH PRESSURE COOLANT INJECTION SYSTEM (IE41)

CORE SPRAY SYSTEM (1E?!)

i

N G'-W-2 -151-2 1E51+MOV 032 b1A-2 a 1E41*MOV 032 fy y X-15 "l ,. -1E41-16"-W-2-151-2 1E41 -4"-W-9-151 -2 X-19 .f. ,

.y.

y v. '- 4 g ,4. .

• %:[q ING)

% g .,..

X-21B ' *:A 4 s W R-246-301-2 1E11-16'i 1E21-14"-WR 21-1514 1E11 MOV 042B .s 1E11-16'!WR-254-301-2

~

1E21-12"-W R 301-2 .

1E11-16'iWR-308-151-2 X-108 L i (-( RHR HEAT EXCHANGER

---1E11 *E-034B D

~

...- -1E11-16'!WR-234-301-2 i

1E21-10"-WR-27-151 2 1E11+MOV 031B ..-C - - -1 Ell-18'iWR-248-301-2 1E11-20'!WR-206 ~1E11-20"-WR-232-151-2 7

• 1E11-20"-WR-212-151-2 .f g - 1E11+MOV 035B 1E11*MOV 031D Q. , ,

IE11-20"-WR-214-151-2  :. /

1 l f WR-224-I2 g.4

'W f.T V

WR-210-1512 SHUTDOWN COOLING ROSS CONNECT LINE "lLIZED FOLLOWING

.)

FIG. 212.107-3 ECCS COMPOSITE PIPING PL AN EL.20'-O TO 40'-O" SHOREHAM NUCLEAR POWER STATION-UNIT 1 FINAL S AFETY ANALYSIS REPORT