ML20151L922

Summary of 880708 Meeting W/Util Re Recent Valve Leaks in Emergency Feedwater (EFW) Sys,Resultant Overtemp of Part of Emergency Feedwater Sys & Other Components & Corrective Actions Taken & Planned

ML20151L922
Person / Time
Site:	Crystal River
Issue date:	07/22/1988
From:	Silver H Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8808040185
Download: ML20151L922 (20)

Text

. $ f e .

July 22, 1988 Docket No. 50-302 DISTRIBUTION l cq U5HHTTTe ACRS (10)

LICENSEE: Florida Power Corporation NRC & Local PDRs PD22 Reading FACILITY: Crystal River Unit 3 H. Berkow H. Sil.ver

SUMMARY

OF JULY 8,1988 HEETING ON OGC CRYSTAL RIVER 3 (CR-3) EFW LCAKS E. Jordan B. Grimes NRC Participants On July 8,1988, representatives of Florida Power Corporation (FPC) met with l members of the NRC staff to discuss the recent valve leaks in the emergency  !

feedwater (EFW) system, the resultant overtemperature of part of the EFW system and other components, and corrective measures taken and planned. The list of attendees, agenda of the meeting, and slides presented during the meeting are enclosed. Also enclosed is FPC's letter dated July 14, 1988 which incluces a sumary of the July 8 presentation. (The information transmitted i l

with that letter will be reviewed separately.) l Significant matters of concern identified at the meeting and not discussed in the enclosures are as follows. These matters will be pursued during the review l of the FPC letter of July 14, 1988 or in future discussions. l

1. The effect of multiple heatups and cocidowns experienced by the EFW system before and during the current events has not been examined.

2. The effect of cooling a hot system by injection of cold water at low flow rates, as was done at CR-3, has not been examined. Compone'nt darage has occurred at other plants as a result of such cooling.

3. Operability of the valves in cuestion at full EFW flow has not been positively demonstrated. FPC stated this would be done during the cutage to repair the valves. l

4. It was agreed that the matter of appropriate testing of the valves in l question as part of the Inservice Testing (IST) program would be discussed at a future meeting.

5. FPC noted that consideration is being given to replacing the EFW contain-ment isolation valves FWV-43 and 44. The NRC staff noted that these valves should be "leak tight", as in many other plants, to minimize the likelihood of water harr.er and other problems, p\

8808040185 880722 PDR ADOCK 05000302 p PNU

r , ,

2_

Members of the NRC staff concluded that as a result of this meeting there was no reason to change conclusions reached in its Confirmatory Action Letter dated July 6, 1988.

Sincerely, Harley Silver, Senior Project Marager Project Directorate 11-2 Division of Reactor Projects-1/II Office of Nuclear Reactor Regulation

Enclosure:

As stated cc w/ enclosure:

See next page L 29h PM:P 11-2 D:P 2 DtMled FSilver:bg HB w 7/FY88 7/$/88 7/>)/88

1

, , , 1 l

Mr. W. S. Wilgus Crystal River Unit No. 3 Huclear  !

Florida Power Corporation Generating Plant l

cc: l 1

Mr. R. W. Ne15er State Planning end Development  !

Senior Vice President Clearinghouse and General Counsel Office of Flanning and Budget Florida Power Corporation Executive Office of the Governor P. O. Box 14042 The Capitol Building l

St. Petersburg, Florida 33733 Tallahassee, Flcrida 32301  ;

1 Mr. P. F. McKee Mr. F. Alex Griffin, Chairman {

Director. Nuclear Plant Operations Board of County Comusioners Florida Power Corporatior, i Citrus County P. O. Box 219 110 North Apopka Avenue )

i Crystal River, Florida 32629 Inverness, Florida 36250 '

i Mr. Robert B. Borsum Mr. E. C. Simpson Babcock & Wilcox Director, Nuclear Site Nuclear Power Generation Division Florida Power Corporation Support 1700 Rockville Pike, Suite 525 P.O. Box 219 l Rockville, Maryland 20852 Crystal River, Florida 32629 Resident Inspector U.S. Nuclear Regulatory Comission l 15760 West Powerline Street Crystal River, Florida 32629 Regional Administrator, Region II U.S. Nuclear Regulatory Comission 101 Marietta Street fl.W., Suite 3100 Atlanta, Georgia 30323 Jacob Daniel Nash Office of Radiation Control Department of Health and Rehbbilitative Services 1317 Winewood Blvd.

Tallahassee, Florida 32399-0700 Administrator Department of Environmental Regulation Power Plant Siting Section State of Florida 2600 Blair Stone Road Tallahassee, Florida 32301 Attorney General Department of Legal Affairs The Capitol Tallahassee, Florida 32304

ENCLOSURE MEETliiG WITH FPC ON 5/8/88 l EFW CHECK VALVE LEAKS - SYSTEM OVERTEMPERATURE i l

1 NAME ORGANIZATION Harley Silver NRR CR-3 PH I Rolf Widell Director, Nuc. Ops. Site Support, FPC Walter Wilgus Vice President, Nuc. Ops. - FPC Vincent Roppel Mgr, Maintenance and Outages - FPC Jeffrey Tedrow Resident Inspector, NRR I Bruce Wilson Chief, Projects Branch 2, Reg.I l Herbert Berkow Director PDII-2, NRR l Ashok THadani DEST /flRR l Alan Ferdt Chief, Engineering Branch, Reg. II John Schiffgens PDII-2/NRR ,

John Craig Chief,SPLB/ DEST /NRR Jerry Wermiel Sectior. Chief SPLB/ DEST /NRR Jim Richardson EAD/NRR/NRC L. B. Marsh liEB/NRR/flRC Ted Sullivan MEB/NRR/NRC i Gary Jackson Gilbert / Commonwealth - Engineer Daniel Biss Gilbert / Commonwealth - Structural Engineer Robert Vaughn Gilbert / Commonwealth - PM CR-3 Ronald Fuller Sr. Nuclear Licensing Engineer, FPC Frank Fusick Sr. Nuclear Engineering Supervisor, FPC Mike Clary liechanical Engineer, FPC Gary Becker Manager, Site Nuclear Engineer Services, FPC l Steven Varga 0:RP/NRR Ray Wittman FPC i

t i

JULY 8, 1988 MEETING WITH NRC ON EMERGENCY FEEDWATER LEAKS I. Introduction (W. Wilgus)

A. FPC Staff Introduction B. Purpose of Meeting C. Format of Presentation o Current Event Description (Refuel VI to now) o System Description o Previous Experience (Prior to Refuel VI) o Future Actions D. Opening Statement II. System Description and Problem Definition (R. Whittman)

A. Emergency Feedwater System Overview B. Emergency Feedwater Line Heatup/ Leak III. Current Event Description conditions and activities associated with the emergency feedwater line heatup from Refuel VI to present.

A. Chronological Event Description (R. Widell)

- Refuel VI outage begins on Sept. 15, 1987

- Repair of valve EFV-33 and inspection of ETV-15, 16, 17, 18 and during Refuel VI

- EFV-18 and EFV-33 identified as leaking to a minor degree on January 7, 1988

- Startup from Refuel VI on January 10, 1988

- Plant trip in March 1988 resulted in a Mode 3 outage

- EFW line heatup during startup from March outage, Engineering requested to track and review problem (March 9, 1988)

- Warm EFW line discovered on May 27, 1988

- Nonconforming Operauions Report generated on June 19 due to excessive heatup of EFW line (NCOR 88-81)

, - Visual inspections by Engineering performed l beginning on June 20 (additional inspections conducted on June 21 and 22)

- One hour report made on June 21, 1988 based on exceeding the EFW line temperature design specifications

- Pressure test (Type B) of penetration #109 on June 21

- Piping restraint discovered damaged on June 21 during visual inspection

- Analyses initiated for the following on June 21:

Thermal stress on penetration steel Review of effects on penetration concrete Thermal analyses of EFW piping inside RB

- Temporary modification installed to short circuit leak at EFV-18 on June 22

- Piping restraint EFH 126 repaired on June 23

- Thermal analysis of EFW piping outside RB initiated on June 23

- Radiograph of FWV-43 performed on June 23

- On line leak repair of EFV-18 on June 28 B. Direct Actions (G. Becker)

1. Analytical 'J valuations

a. Penetration Analysis o Concrete o End plates and sleeves

b. Thermal Flexibility Analysis of Piping  !

o Penetration 109 to OTSG-B l o EFP-2 to Penetration 109 ll o Penetration 424 to OTSG-A

[

1

c. Pipe Support Evaluations l

o EFH-126 o Supports in R.B.

d. Environmental Qualification of EFV-55

2. Inspections and Tests

a. Field walkdowns of piping and supports in Intermediate Building

b. Walkdown of piping and supports inside R.B. l

c. Visual inspection of Penetration 109

d. Pressure test of Penetration 109

e. P.T. Weld between 3/4" plate and process pipe

f. Radiograph FWV-43 l

3. Temporary Modifications

a. Injection System o Description of system o Design and seismic considerations o How system performed

b. Injectable Sealant in EFV-18 o Description of injection technique o Pressure stress calculations o Other considerations l l

IV. Previous Experience (V. Roppel)

A. Summary of Previous Experience l

B. Maintenance History V. Future Actions (V. Roppel)

A. Conduct a repair outage for EFV-18 on or before October 15, 1988 o Repair EFV-18 and EFV-33 o Inspections of EFV-16 and FWV-43 o Review conditions of above valves to determine if inspection of "A*' OTSG flowpath valves is needed B. Maintain PM program on check valves in the EFV flovpath valves addressed herein.

C. Followup to EPRI Check Valve Applicability Guideline Review for appropriate maintenance ii.spections.-

D. Review upgrade of the emergency feedwater piping, components and penetrations to accomodate expected pipe temperatures.

V. Questions and Answers ( All )

a.

l == ==

~

~ FLOWPATH FOR EMERGENCY FEEDWATER ,

I ~

-- _$_2__ ,-.

sesv-cs a sesv-e6 a senes j

h/ ruwA10t ,

I , b A'Y 'P spp.eas sev-est C FW-# ""

1 ro  ;; 'L j ) 1 37EAss mar - #

. r D1-2 13 AmagegK 5

t M Aav-G 4 1

1 H

SB W hg l8 g i i <.

[ snMe

_.w..- grv es ,

l

• • 1 u ,

a

== = -d5 :h:

( f. ,i 3 FW-M FW-@4 FW-e3 g

! l E sc ,I f

1 m' r' %

t. . E j i l

. 1 ' e v-ee 16 + ,

yy er BC e W E 4 t ;t ,, i P

>m '

J k

. rw-gr 6 20 000 CAP '

l 150.000 NIN

d g '

L T 1 ~H w l

• n r,

2s i N1m ,

l i f Mg g .

I 200.000 CAP '

./

eI 1

r.

- ^~.

m -

sv-F g ,

\

i casassess N

[ s a nnsE 4 ,-

l #3 ..

I

• av.es yve ,

ac o

• N l

l l u

rwv-en MG !.ONGER avv-es tw-44 l^

i T 3. '"

CONTROI.i.ED M

'l cw-e h- - ,

en se IS av-ee ac e .v4P W L , . . -

--- ~~ ~

REFUEL VI OUTAGE BEGINS ON SEPTEMBER 15, 1987 REPAIR OF VALVE EFV-33 AND INSPECTION OF EFV-15,16,17,18 & DURING REFUEL VI EFV-18 AND EFV-33 IDENTIFIED AS LEAKING ON JANUARY 7, 1988 STARTUP FROM REFUEL VI ON JANUARY 10, 1988 PLANT TRIP IN MARCH 1988 RESULTED IN A MODE-3 OUTAGE EFW LINE HEATUP DURING STARTUP FROM MARCH OUTAGE WARM EFW LINE DISCOVERED ON MAY 27, 1988 NONCONFORMING OPERATIONS REPORT GENERATED ON JUNE 19 DUE TO EXCESSIVE HEATUP 0F EFW LINE (NCOR 88-81)

VISUAL INSPECTIONS BY ENGINEERING PERFORMED BEGINNING ON JUNE 20 ADDITIONAL INSPECTIONS CONDUCTED ON JUNE 21 AND 22)

ONE HOUR REPORT MADE ON JUNE 21, 1988 BASED ON EXCEEDING THE EFW LINE TEMPERATURE DESIGN SPECIFICATIONS PRESSURE TEST (TYPE B) 0F PENETRATION #109 CN JUNE 21 PIPING RESTRAINT DISCOVERED DAMAGED ON JUNE 21 DURING VISUAL INSPECTION

e' i

l l

l l

l ANALYSES INITIATED FOR THE FOLLOWING ON JUNE 21: l o THERMAL STRESS ON PENETRATION STEEL o REVIEW 0F EFFECTS ON PENETRATION CONCRETE o THERMAL ANALYSES OF EFW PIPING INSIDE RB TEMPORARY MODIFICATION INSTALLED TO SHORT CIRCUIT LEAK AT EFV-18 ON JUNE 22 EFW LINE RETURNED TO NORMAL TEMPERATURES AND PIPING RESTRAINT EFH-126 REPAIRED OH JUNE 23 RADIOGRAPH OF FWV-43 PERFORMED ON JUNE 23 ON LINE LEAK REPAIR OF EFV-18 ON JUNE 28

0 0 C t.

p.t Af p4'0 gW f,g99,,%,$.Q"6' g .< ,;

,,, g s'

\

< H l

4 gpp l**

l -

w l

\ ?_'o* _ . ,

l Y g pr' 0 R/6/O SPR/HG

\

4Gruees-uw g AMcHM

t

, t,

%"H/au.3 jy,,,,,'.;.- <

7 . ,.:., . ..

', 4 4 i,,.,,,s ' ;,s

. .s

,t 14'& l .'e-<4<l  ? I '

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

h"g t y ss/1Bwn$s 'YY! $ l'

, i ,. g i h, i! !! d i ie I

il l l i

i il I; lil 11 Igil I

il I l I

' il l;i il l

i i

,i ll ii l ir a 6'$ . .. .

-i".",."..;...

r '- ..".

"z.

a ..

w ".

c . , , ,e ; .-

IN rEAMEDATE ' ' * .: . '.' ' 3. . .

. , ' * .s.?

REACTCA BulLCIWC3 . .' ,I ' '

. h. .

B UIL OfM G i,.'f,f

.S'. '

.. .'~*

\

t k

\

\f

- _. . . _ . - - . _ _ , , _ _ _ . _ _ _ _ m ~ _ . _ _ _ ., . . . _

e e II d

! I.X- 'tp X.

• XI o i H

. () ..

- !X!

7 ,,

h

! l>

o% n

'1 <  !,

[ T-~

4 3

!

• T 11 l-F lllm l

=

s t

.I Y.

s k \

I ..

% .0

~

l j% \

~I

.'\

i is ,

k Q

1 s=

( I (k

,S S 2

I

[ \

\

\. .

\

')

,o .

\

., r.t s o

4  :

\

h

<j. s. ,

fl .,

\

( #

s

{

4 4 ()

/

. g

\w k e

q Ya g W

_S 5

MAINTENANCE HISTORY .

t..

DAIf PROBLEM CORRECTIVE MAINTENANCE l

i FWV-43 4/12/79 ~'TCK VALVE BACKLEVAGE DISASSEMLE Alm REFUR8ISH l 9/14/79 BONNET LEAK REPAIRED NINGE PIN SEAL PLUG LEAK 1/5/83 CNECK VALVE BACKLEAKAGE .DISASSENBLE AND REFURBISH 3/15/83 CNBCK VALVE BACKLEAKAGE DISASSENBLE AND REFURBISH 1/30/84 PLagEB PN (50ER 84-3) DISASSENBLE AND REFURBISH EFV-18 6/19/90 INSPECT DISASSENBLE AND REFURBISH 6/20/83 BACKLEAKAGE DISASSEfBLE AND REFURBISH 1/30/84 PLAINIED PM (50ER 84-3) DISA5SEISLE AND REFURBISH 12/3/87 PLAfeIED PM (50ER 84-3) DIIA1583LE AND REFURBISH 1/26/88 BONNET LEAK (SEAL RING) T M NIGNER - NOT SUCCESSFUL a EFV-16 6/15/80 INSVECT prueesmiF AND REFURSISH 4/25/84 PLANNED PN (50ER 84-3) SEM555BLE AND REFURRISN

12/5/87 PLANNED PN (50ER 84-3) DISA55DSLE AND REFUR8ISN p

EFV-33 10/23/85 BONNET SEAL RING LEAK RETOROUED STUD NUTS 12/2/87 BODY TO BONNET REPLACED SEAL RING i

3/23/88 BODY TO BONNET RETORQUED - NOT SbCCESSFUL a

j EFV-14 12/3/87 BODY TO BONNET DISASSENBLE & REFURBISH - SEAL RING

! REPLACED

( '

SUPMARY OF PREVIOUS EXPERIENCE DAIE PROBLEN CAUSE LER's 1/19/83 82-076 HEAT FAILURE OF FLOW TRANSDUCER FW-43 LEAK 8/19/83 83-029 FAILURE OF FLOW TRANSDUCER HIGH HFAT FAILURE OF SEALANT MOUNTING 8/24/84 83-043 FAILURE OF Fi.0W TRANSDCCER FW-43 LEAK NCOR's 4/24/84 84-101 HEAT FAILURE OF FT FW-43 LEAK 6/18/84 84-148 FAILURE OF SPAN ACCURACY FW-43 LEAK TEST (PM-243) i UDIR i

! 10/15/82 82-25 POST TRIP FAILURE OF FW-39 17 MIN FEED TO HIGH N0ZZLE FOLLOWING RESTART OF MFP I

l l .

a - .- -s -

$ I O l

ees g psl gh=$ h k I a- ,

\\ ,

a

...---....y R

e Q .

g n

Q .

0

-- - ~ - ---

- - , , , , . . _ - , ,w-. _ . . - - --...--,e,,,, __-_ , , _ , _ , _ _ _ _ _ , - _ _ _ _ _ _ _ _ _ , , _ _ _ _ _ _ _ . _ _ _ _ , _ _ _ _ _ _ _ _ _ -

l EUTURE ACTIQHS A. REPAIR OUTAGE IN OCTOBER 1988

1. FWV-43: REPAIR BACKLEAKAGE 1 e DISASSEMBLE AND INSPECT VALVE INTERNALS l e REPLACE WORN PARTS, LAP VALVE DISK SEATING SURFACES AND CHECK FOR PROPER l SEAT / DISK ENGAGEMENT e POSSIBLE MACHINING 0F INTERNALS SURFACES OR EVEN VALVE REPLACEMENT '

2. EFV-18: REPAIR SEAL RING LEAK e DISASSEMBLE AND INSPECT VALVE INTERNALS e REPLACE ALL WORN PARTS, LAP VALVE AND DISK SEATING SURFACE AND CHECK FOR l PROPER SEATING l e POSSIBLE WELD BUILDUP AND MACHINING 0F SEAL RING SEATING SURFACE, OR VALVE REPLACEMENT

3. EFV-16: REPAIR BACKLEAKAGE e DISASSEMBLE, INSPECT, AND REPAIR

4. EFV-33: REPAIR SEAL RING LEAK e SAME EFFORTS AS EFV-18

5. FWV-44: POSSIBLE DISASSEMBLE AND IN"02CT e EVALUATc/ CONFIRM BACKLEAKAGE

FUTURE ACTLOSS. CONT'D B. PM PROGRAM ON EFW CHECK VALVES

1. INP0 SOER 84-3 (STEAM BINDING OF AFW PUMPS FROM CHECK VALVE BACKLEAKAGE  ;

o SHIFTLY CHECK OF EFW PUMPS AND DISCHARGE l PIPING  ;

o REFUELING INTERVAL INSPECTIONS OF EFW CHECK VALVES

2. INP0 SOER 86-3 (CHECK VALVE FAILURES / DEGRADATION) o ENGINEERING REVIEW 0F CHECK VALVE APPLICATIONS (SIZE, TYPE, ORIENTATION)

IS COMPLETE o RECOMMENDATIONS FOR POSSIBLE CHANGES TO PM PROGRAM TO BE EVALUATED BY ENGINEERING AND MAINTENANCE

3. COMPLETE EVALUATION OF MAINTENANCE HISTORY C. COMPLETE EVALUATION / UPGRADE OF EFW PIPING FOR EXPECTED TEMPERATURES

1. CONTAINMENT TEMPERATURE EFFECTS

2. PIPING INSULATION EFFECTS

3. ACTUAL DESIGN TEMPERATURE LIMITS

t .

3

- Fi+'s t.

l Florida Power COaPOaATQN July 14, 1988 370788-12 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 )

Subject:

Crystal River Unit 3 Docket No. 50-302 Operating License No. DPR-72 Emergency Feedwater <

Dear Sir:

Florida Power Corporation (FPC) has reviewed the Confirmation of Action Letter dated July 6, 1988. The attached information is provided as a supplement to the correspondence 1988, concerning the emergency feedwater leaks and summarizes the submitted on June 28, 1 July 8, 1988 presentation in Rockville, Maryland. The issues associated with the appropriate testing of check valves employed as ,

containment isolation valves, such as FWV-43 and FWV-44, will be l discussed further at a follow-up meeting yet to be scheduled. The issues associated witn oc- corrective actions for previous events were discussed at the July 8 meeting and will be the subject of .

l separate correspondence with Region II associated with Inspection Report 88-18. l l

1 If you have any questions, please contact this office.

Sincerely, l D

'K.R. Wilson, Manager Nuclear Licensing KRW/REF/dhd Attachment xc: Regional Administrator, Region II Senior Resident Inspector

^

GENERAL OFFICE: 3201 Thirty fourth Street South

• P.O. Box 14042

• St. Petersburg, Florida 33733 * (813) 866 5151 A Florida P%ress Compang _ _ _ - .

l 1

azeraerIm oF scrim IzrIsa I SGGERY OF JUIX 8,1988 HESENMFIm  !

Baclazzcund m QZrrerft EFW Line Heat m Evesit:

Durirn Refuel VI, several emergency feedwater (EFW) system valves underwent refurbishira. These efforts included the repair of EFV-33 ard the inspection of EFV's 15,16,17, and 18 (see Attachment 1 flow schematic) . Two (2) of these valves, EFV-18 and EFV-33, were identified as haviry body to bonnet leaks during Refuel VI startup activities on January 7, 1988. Efforts to secure the leaks were unamaful. The unit was returned to service on January 10, 1988. )

I l

Crystal River Unit 3 experienced a 13 actor trip on February 28, 1988, due to a I malfuncticn in the feedwater systen which resulted in an EFW injection (IIR 88-06). 'Ihe plant entered a MXE 3 outage for 4 days to investigate the cause of the trip. The unit was returned to servios cn March 3,1988, but limited to 1 60% power. The unit was removed frca service on March 7 and entered MXE 2 to effect repairs on the main feedwater flow ocntrol vdvs. During the startup '

cn March 8, the Operations staff cLiscovered all four EFW lines hot. 'Ihe check valves in the EFW lines waru apparently allowinJ scas backflow causing the '

heat up of the EFW lines. Estimates of the taperatures ard ccniitions of the , j lines were provided by Curations to Engineering. Follow-up inspections of the# i EFW lines determined that the taperature had returned to normal following a ~

cool down prrv=== which involved flowirq a small amount of energency feedwater '

through the lines. This action evidently caused the leaking check valves in the lines to reseat since these lines remained cool until May. l On May 27, 1988, during a routine plant walk down, the Shift Operations Technical Advisor (SCTIA) die'evered the arc rgency feedwater pipirq near flow ,,

control valve EEV-55 to be warmer than normal, iniicating check valve FWV-43 lj was not preventing backflow into the energency feedwater line frun the steam '

generator. This condition was dimaad with the Shift Supervisor on duty and incraaamd awareness of the condition was maintained by site perscnnel. By June 19, the taperature of the line frce the reactor buildirq penetration to EEV-18 was hot enough to cause scan flashing of the water leaking frun EEV-18. A lbn- )

Conformirq Operations Report (lKOR 88-81) Vas written cm June 19, and upon further investigaticn a cne hour report was initiated on June 21. The report was hamad on armading the design basis tenperature for the emergency feedwater line. Prior to this time, the line was thatqht to have been designed and analyzed for nu:tt higher taperatures.

As a result of this heat up, visual inspecticas and engineering analyses were conducted. 'Ihase activities incitded:

o Vi a m1 inspecticms of the penetration and walkdowns of accessible piping ard supports by Engineering perscnnel o Pressure test of Penetration #109 !l ll o Dye perscant i,Wicn of high streau weld on Penetration #109 l o Thermal stress analysis cn penetraticn steel l 1

1 i

o Review of ta perature effects on penetration ccncrete o Thermal analyses of EFW piping inside and outside of reactor bui1 ding o Installation of a teperary codification to reinject leakage frm EEV-18 to stop the EEW line heat up, o Reevaluation of str-a caused by seismic, deadweight ard internal pressure forces as a result of the high taperature ceniitim.

o Radiograph of check valve FW-43  !

o Repair of pipe restraint EEE-126 o Leak repsir of EFV-18 l l

l Details of these activities were provided during teleconferences and in a j letter to the NRC dated June 28, 1988. Additicnal information ccreerning the penetration analyses, the thermal flexibility analyses of the piping, the pipe support evaluations, and the leak repair of EEV-18 is provided below.

I% = ualicn Analyses:

S.

The per Lcaticn acosists of a 14" sleeve and a 2" thick and plate inside the -

reactor buildirg. A 6" ScNrhtle 160 pr-a pipe passes through the center of the sleeve and end plate. Welded lugs aamru the sleeve to the wm & te containnent wall. A 3/4" plate is welded cn the sleeve to provide a closure plate to facilitate testirg of the penetration. The 3/4" plate is r2 considered part of the containent bourxiary, and is not credited as a pipe support in the piping analysis.

A two dimansional, steady state heat transfer nodel was developed to determine the maHun anticipated wn=te taperature attained near the par-traticn inside the Raactor Building (RB) . Besed on a mav4== process pipe taperature (measured) of 480* F, a mavim= wc=i.a tenperature of 214 *F was predicted. .

The mMal was validated by ccmparing its predications with measured values )

obtained at a locaticn above the penetraticn cutside the RB. Excellent I agt=mant was fourri (mavin = measured was 195'F) .

Although the overall w 4=te design tape.rature for CR-3 is 150*F(haaad on ACI 318-1963), FPC considers the criteria established in ACI 349 & 359 to be relevant for this transient situation. ACI 349/359, Appendix A, permits localized ocncreta taperatures adjacent to per=Lrations to be 200'F for long term periods (ocntimous) and 350'F for short term erergency situaticos, Thereform, FPC concluded and visual impaction confilmud the orrete was not adversely affected by *his elevated taperature condition.

The penetratico sleeve and end plates weis analyzed by conservatively e==ing a 300'F differsntial ta perature betw en the sleeve and procass pipe. This ev= dad the actual maamited differential taperature of 208'F. Undar a 300'F diffarential ta perature, it was determined that the 3/4" closure plate would be stressed beycod its elastic limit (28,000 pounds-force) . The cou. sponiire stress in the end plate frm this load b 6,000 psi. This stress, when added to the stress frun the piping loads of 5,000 pai, are well within the allowable stress of 45,000 psi, thereby assuring that ccntalnnent integrity was never canptruiami.

2 _ , _. . _ _ _ _ . ___

The 3/4" closure plate was loaded beyond its elastic range and was perfouning in the plastic range, however, FPC does not consider this to be a problem since its only fanctico is to facilitate tasting of the per-hation. It has not been taken credit for in maintainhg containment integrity or as a pipe support.

An additional inspection and dye penetrant test were conducted to de .-trate no degradation has occurred to the 3/4" plate.

Pining Analysis:

A thermal flexibility analysis was performal en the piping inside and outside of the Reactor Btilding (RB) . The analysis was pedu i at the elevated temperatures.

The piping inside the RB was analyzed at a tanparature of 480'F frun Fenetration #109 to a point midway to the steam generator, and at 590*F for the remaining portion. This taperature profile is a ocomervative --n= tion since the actual gradient between the steam generator and the p.-t.ratice would be more gentle. This portion of the stzuss analysis d <.t. rated that the pipirg inside the RB frcan Penetration #109 to the steam generator was not stznamari above ANSI B31.1,1967 code allowables during the elevated "anparaturs conditicn. Mav4== stress in the piping was calculated to be 17,256 psi which -

is balcw the allowable of 22,500 pai. S-A thermal analysis was performed cm the piping cutside the RB at the elevated temperatures maamtred frcan EFV-18 to Penetration #109. The system supports were modeled to reflect the as found ecodition based on the visual inspections.

The results of the analysis iniicated that the piping outside the RB was not subjected to stresses in excess of code allowables. The nav4== calculated thermal pipe stress was 14,427 psi which was less than the 22,500 psi allowable.

A pressure, deadweight, and seismic (SSE) load analysis of the piping outside the RB was performed with EFH-126 mieled in the <4anatyal conditice ani providirg only translaticnal and torsional restraint. Under these load combimtions, the mavi== calculated pipe stress was 16,445 psi which was less than the allowable 18,000 psi.

Pine n==v=t EvalueHr=1s:

The effects of the loads inrraad by the piping on the pipe suoports were evaluated. The pi stresses in am===pe supports inside the reactor building were not subjected to of 331.1 code allowables durirq the elevated tamperature cxxxiiticn. In additicm, only one (1) pipe support cm the subject piping inside the RB utilized andbor bolts. Under the worst case acabined loading, the anchor bolt safety factor was redd to 5.6 ocspared to the mininza 4.0 specified in the NRC Bulletin 79-02. Therefore, pipe sqports inside the RB remained fully functicmal during the elevated tamparaturu coniition.

The supports on the piping outside the RB were also evaluated. Base plate stressas in two supports (EFH-129,130) would have swearied B31.1 allowables during sei.enic events. The worst case value of 33,100 was balcw the yield sL=Wth of 36,000 psi during this postulated loading conditicn.

3

l l

Durirq the postulated SSE loading coniitico, anchor bolt safety factors for two supports would have been below the mininnn of 2.0 as specified in Bulletin 79- l 02 . The minimum calculated safety factor was 1.7 at EFH-132, which is above l 1.0 ard does not constitute a failure of the anchorage. It is inportant to I enphasize that these are worst case load cxznbinations based on seismic, deadweight and themal pipe loads. Actual loading during the elevated tarperature conditions, excluding seismic, was within allowables at all times.

EFV-18 Isak Ds- Iv:

On June 23, 1988, a temporary modification was made on EEV-18 to seal the body l to bonnet leak. A low temperature, flexible sealant was injected by means of a manually operated pressure gun into the spaces en top of the bonnet and inside l

the lock ring of EFV-18 which totally eliminated the leak. A diagram of EFV-18 is prwided in Attachment 2. The sealant was injected into the valve first l I

throtqh the modified cap bolts to seal the 1. *.erior space and bolt screw holes (area "A"), and seccrid through the two valve vent ports to seal the spaces in {

the valve lock ring and cap (area "B"). The sealant was injected at pressures '

below the systan pressure to preclude sealant frm entering the EFW line except for the final few strokes of the gun. 'Ihese final scrokes nust emm at an .  !

injecticri pressure ec= Ming system pressure in order to ocapress the sealant.. l and stcp tha leak. This method minimizes the retential for sealant to enter '

the valve internals. An EFW system flow test of approximately 50 gym was used S to ocnfim valve functicn was unaffected by the use of the sealant. A test of .- .l the valve integrity was conducted by placing the EfW system into recirculation thereby, generatirg maximum operating pressure. The chaunistry of the sealant '

was certified to be within acceptable values and crzrpatible with valve material. I Previous W ""lerloe with EFW Lirm Heat m:

A review of previous EFW events revealed three (3) License Event Reports (IER's) and two (2) Nonconfoming Operaticos Reporta (NCOR) had been written documenting the problems with the flow tram %rs on the EEW lines due to high temperatures. During the time frame of these events, FW-43 was sepacted of backleakage which resulted in the EEW line heat up. The flow paths for the leakage were not identified in the derxments. A sumary of previous events is provided in Attachment 3.

Planned inspecticos of EFW check valves were performed during cutages in 1980, 1985 and 1987. The 1985 and 1987 inspecticos were planned EM's instituted as a result of SCER 84-3. A review of the Laintenance history of the E!W systen valves was performed ard FW-43 and EFV-33 were identified as having a t end of pcor perfomance frta the standpoint of backleakage and leaks to atmosphere.

Evaluaticris have been performed in accordance with the Institute of Nuclear Power Operations (INPO), Significant Operating Event Report (SOER) 86-3, which also identified these valves as cardichtes for additional inspecticn and

• W.

An anncunced NRC inspecticm was conducted at CR-3 durirq October 5-9 1987, to determine the current status of check valve test pr,an. as a suoset of the industry as a whole and to determine the rspmisiveness of the industry to the INTO r+1--ndaticns for improving check valve testing pre aam (SOER 86-3).

During the irWien, the NRC identified EEV-18 and EW-43 as having maintenance prtblers and as being in::luded in the SOER respese program.

4 . _ _ . - .__ -

i l l

l l

P\1ture Action:

Florida Power Corporation on, or before, Octe*vn- 15, 1988, Will permanently repair EEV-18 and EFV-33 body to bonnet leaks, and will inspect and repair, as many, IW-43 and EFV-16 seat leakage. In addition, FW-44 will be inspected and repaired, as rweey, based on the results of the inspections and rgair efforts of FW-43. FPC is currently evaluating the appropriate post maintenance test and acceptance criteria for EW-43 and FW-44.

Hourly temperature readings of the EFW pr-= pipe through Penetraticn #109 will be maintained until a permanent fix is made. 'Ihe onca per shift check of EFW punps discharge piping will continue, as well as refueliM interval inspecticos of EfW check valves. A cmplete evaluatico of the maintenance history cn these valves will be performed and utilizei in conjurction with the INPO SOER 86-3 evaluations to establish r= - ndations for possible changes to preventative maintenance prupams.

A review will be performed to upgrade the EFW piping for actual taperatures expected in the syste. 'Ihis review will include high energy line, stratification and water ha::rer offects as appropriate.

i l

l 1

l l

l I

e 5

03E'IHRTICN CF ACI'ICN IJ! TIER RESPGM Daarrruses to Soecific Ctmfiratico of Actim Istter I*:

Item:

A diemaion of the effect of potential DV systan degradation en the reliability of the systen and on long-term plant safety.

T w -=:

The primary adverse consequence was DV reliability. Florida Power Cbrporation maintains that the DV reliability was traffected since the piping, penetration and supports were decretratai through analyses to be capable of performing their intended functions within the design allowables excluding seismic loads.

The piping has been returned to within the design tar @erature and will be maintained within this specification, thereby, alleviating the adriitional thermal stresses experienced during the abnormal heat up events. The valves and systaa performance were and continue to be unaffected by the recirullation systen or the injecticn sealant as di aai in earlier con +4ence and the presentaticn sumary contained herein.

Itse:

A hiption of the repair of DV-33 and of the injected sealant repair of DV-18, results of the repairs, and description and results of testing to verify u.mprent/systen operability after repairs.

wae:

The co-line repair of DV-33 is under evaluatico. Imakage frcan DV-33 has decreased and will ccotinue to be mcnitored. The injected sealant repair of DV-18 and ==*=qmnt results and tests are dieW in detail in the previous pages as part of the July 8, presentation smmary, Itaa:

The results of all thermal and stress analyses of all involved piping, penetraticos, and sumorts.

Response

The esults of the thermal and stress analyses of the penetration, piping and supports are diew in the July 8, presentatico sumary. Additional analytical informaticn is provided _n Attachment 4.

Itan:

Evaluaticn of the safety inplicaticos of the valve leakage and DV systan overtamperature and of the DW systen with the interim fix and the taiporary repairs in placm.

6 _ __ _ _ . _ . _ . _ _ _ _ . _ . _ _ _ _ _ . _ . _ _

'

• c, i

l 1

l Response: 1 The EEV systen valve leakage did not pose any safety m. The leak rata did not aw=d 0.9 gpn and had no adverse effects m the EfW flew to the steam generators. The potential for water hamer was considered unaffected by these events due to the following:

o The piping diameter was small o The piping has short horizmtal runs and manarcus ben:s o The piping has rigid supports at the majority of the pipe bends o The EEW enters into the steam space in the steam generator o the majority of the piping remained filled with water because the tamparature was below the satulation temperature of 540'F for a majority of th piping.

o The pipinJ ger arally rises frt:st the EFW pays to the steam generator..

The potential for stamm binding of the DW pays was also cmsidered.

Manitoring was initially conducted on a once per chift basis and later S '

%.isi to hourly to quickly identify conditions whicit could lead to -

stasm binding. There was never a problem or threat of steam binding of ths EFW pups as a result of the leaks since other valves in the lines {

were preventing back flow and heat up of the system. The FWV-43 back- j leakage was not ocmsidered to be a cignificant safety ocncern since the '

valve was considered cperable for EPW injection and minor backleakage was 1 of little ocosaquence as lcog as the EfW piping and pays runnined cool.

The cuitainment integrity issue was not considered significant since the {

1 piping and steam generstor prtwided syste integrity in the RB (closed <

systaat inside containnent) . Pbr a steen generator tube npture event, the l

off sita dose caneequences remain well within design basis for this 1 accident.

}

The tamporary injection syst m was carefully designed to ocnsider EfW syntan isolatico, seismic restraint armi relief valve needs in order to adequately protect the reliability of the EFW systsa. The temporary EFW-18 leak repair was d*-M in the July 8 presentation anunmary.

Itam:

Verification that EPW otspcments have not been damarymd by overtamparature events and can reliably perform their safety functicms.

Response

the only adverse coneoquences not addreseed elsewhere involved EEV-55 and EFV-

57. The effects of hiWar than egocted tamparatures on the qualification life of EEV-55 and EPV-57 were evaluated, and sufficient margin is retained for l ocr(cinued operation. Early repair /replar-ent will be evaluated as part of our ,

normal EQ maintenance program.

t 7

. _ _ _ . . - -_ _ _ ._ . _ _ _ _ ______ _ _ _ _ _ _ _ __U

I l

l i

l l

Itaus:

A dieumaion of the possible need for a nore emprehensive inspection and repair prw&-a had cal an evaluation of Current and past valve leakges. l Beepcrise:

Florida PcWer Cbrporation is contimirg to evaluate and develop a rure c +rdersive inspecticri and repair prwsma for <: heck valves. FPC will factor

{

the EFW events into that program. These events do Det ocntradict the  ;

priorities identified by that program (e.g. these were all high priority valves l for enhanced efforts). FPC's detailed respcrise to previous check valve leakage l events will be diet ==ai ucIn thrtughly in separate wri+-dence.

1 I

l l

l

?e G I 9

8

ATTACillENT 1 '

== p FLOWPATH FOR EMERGENCY FEEDWATER

=

v

g. -

-- As ,w_.

5/ FREWHIt T5 b AC r ul' N .svist g, Tet

' I sugase map j

j j i If T 2 -p "

8 M Ane m /,

g l

! l e* *'** .

ep aswee m m .594_. :e:

i

o , -- - jib

,w- ,w.

w

,w_

t j  %= ^

. T g

Gy -s ** a SW ,,

. - , :e:

yv 6 a.v-w

{e at e e d

q, q r 4 k l LA tw-er 200,000 CAP vera s i

' ' i 150.000 MIN

-q 2

y l

" c=> ,_.

• A; m

ik l'

f'i . -

_ ,l r _. _

J_w g'. h 200,000 ( Al' -

Sv-F N EE

- _ i R,

SM W , M Tm m grweg 8

" gg e ,

,h

N4

I e sNGl.h awas tw-.s rww i

M

&8 r41NTiedi i f.D

{

!?

h .

y n 15 a v-en Km C de t w ,

Y w

J 4

F $. $

q 9- _ r b

I4'b 4 .

f e

v 7 1 I

3 6*e x- /

I

% l I

,. 44 if s' s

n g il;i y

Y  ? ' }>  ; '\

~

t g' ,

w ,

a '

W I 4

=

e 4

2 W

I .

R J ~

tn 6 3 1 2

n .

i b.,

2 w 4 C 0 $

W W r Y

_ ._ _. . . - _ - - - - -. -w-_-_____________

ATTACl#ENT 3 '

SUMMARY

OF PREVIOUS EXPERIENCE

DATE PROBLEM CAUSE

! LER's l 1/19183 82-076 'lEAT FAILURE OF FLOW TRAN500CER FW-43 Lt3K j 6/19/83 83-029 FAILURE OF FLOW TRAN500EER HIGH 18 EAT FAILURE OF SEALANT

MOUNTING l 8/24/84 83-043 F4ILURE OF FLOW TRANS00CER FWV-43 LEAK i

J NCOR's 4/24/84 84-101 NEAT FAILURE OF FT FW-43 LEAK

{ 6/18/84 84-148 FAILURE OF SP#i ACCURACY FW-43 LEAK l TEST (PM-243) 1 00ER

]

{ 10/15/82 82-25 17 MIN FEED TO HIGH N0ZZLE POE7 TRIP FAILURE OF FW-39

! FOLLOWING RESTART OF MFP

,i I?

i

\ -

i

~

e d

u. l ATTACHMENT 4 DETAILED ANALYSES i

~ .

EMERGENCY FEEDWATER PIPING f SUPPORT REANALYSIS SUBSEARY INSIDE CONTAINMENT

- At Original Design Temperature of 110 Degrees F All actual stresses on piping and supports were within the allowable limits specified in ANSI B31.1 1967 edition under deadweight, pressure, OBE and SSE loading conditions.

Envelope of maximum worst case pipe stress on system due to deadweight, pressure, and SSE loading conditions = 17,365 psi which is less than the 18,000 psi allowable stress permitted in ANSI B31.1.0 1967 Edition.

All pipe supporc anchor bolt safety factors were within the requirements of NRC Bulletin 79-02. ( i.e. Safety Factor greater than 4 for wedge type anchors under deadweight, pressure and SSE loading conditions. ) Anchor Bolt Safety Factor for EFH-025A ( only support attached to concrete in the effected piping ) under deadweight, pressure, and SSE S-loading conditions = 19 . .

- During Elevated Temperatures All actual stressos on piping and supports were within the allowable limits specified in ANSI B31.1 1967 edition under deadweight, thermal , pressure, OBE and SSE loading conditions.

Maximum worst combination pipe stress due to deadweight, pressure, and SSE loading conditions remained unchanged.  !

Maximum thermal stress in piping = 17, 256 psi which is less than the allowable stress of 22,500 psi permitted in ANSI

, B31.1 1967 Edition.

Anchor Bolt Safety Factor for worst case baseplate for EFH-025A under deadweight, thermal, pressure, and SSE loading conditions = 5.6 .

, _ _ . _ . _ _ _ _ _ _ _ _ _ .13_ - _ _ _ _ . _ _ . _ . _ _ _ _ . ._.

_-H a

OVERVIEW OF PIPING ANALYSIS

( Inside containment )

6" Diameter Emergency Feedwater Piping from Penetration No. i 109 to Steam Generator 3B was analyzed for thermal effects. l Temperatures Used in the Thermal Analysis Included:

- 480 Degrees F from Penetration No. 109 to the approximate

" half-way " point ( i.e. Elbow between supports EFH-8 and EFH-9 ).

- 590 Degrees F from the " half-way 1>oint to and including the Ring Header.

Maximum Thermal Stress & Nozzle Loads

- Maximum thermal stress from the reanalysis was 17,256 psi.

This stress was located on the Ring Header and is -

l acceptable from an allowable stress of 22,500 psi provided ,

in ANSI B31.1.0 1967 Edition and B&W Dw'g. No. 108031, b- '

Rev. O. ' -

Maxisun Primary Stress

- Maximum Deadweight & Pressure Stress = 5,665 pai 0 node 32 Maximum Seismic ( SSE ) = 11,700 psi e node 561 Total = 17,365 psi This stress envelopes the maximum stresses acting on the EFW piping system inside containment. Worst case primary stress is unaffected by thermal transient.

1 l

1_ _14. ," _, _ .._._ _ _. __ . . _ _ .

l EMERGENCY FREDWATER PIPING f SUPPORT REANALYSIS SUt9CARY l OUTSIDE CONTAINMENT

- At Original Design Temperature of 110 Degrees F All actual stresses on piping and supports were within the I allowable limits specified in ANSI B31.1 1967 edition under deadweight, pressure, OBE and SSE loading conditions. ,

Envelope of maximum worst case pipe stress on system due to l deadweight, pressure, and SSE loading conditions = L906 psi, i l

All pipe support anchor bolt safety factors were within the requirements of NRC Bulletin 79-02. ( i.e. Safety Factor greater than 4 for wedge type anchors under deadweight, pressure and SSE loading conditions.

Anchor Bolt Safety Factors ( SSE ) for EFW pipe supports :

Support Safety Factor Support Safety Factor *7 EFH-125 Infinite EFH-133 10.2 -

EFH-126 16.0 EFH-532 (wejits) 4.8 EFH-127 12.0 EFH-542 16.5  !

EFH-128 27.0 EFH-543 11.3 I EFH-129 17.0 EFH-544 9.9 EFH-130 8.7 EFH-545 33.0 EFH-131 14.0 EFH-546 31.0 EFH-132 4.3

- During Elevated Temperature Condition I All actual stresses on piping were within the allowable limits specified in ANSI B31.1 1967 edition under deadweight, thermal, pressure, OBE and SSE loading conditions.

Maximum thermal stress in piping = 1A ,4_27 psi which is less than the allowable stress of 22,500 psi permitted in ANS1 l B31.1 1967 Edition.

Maximum deadweight, pressure, and SSE stress in piping

= 16,445 psi which is less than the allowable stress of 18,000 psi permitted in ANSI B31.1 1967 Edition. EFH-126 was modeled as an X,Z & My restraint only (4 bolts are considered for factor of safety).

Anchor Bolt Safety Factors ( SSE ) for EFW pipe supports:

Support Safety Factor S_upf g Safety Factor i EFH-125 5.5 EFHe133 5.6 EFH-126 7.4 EFH-532 14.5 (maxi)

EFH-127 3.5 EFH-542 4.3 EFH-128 Removed from Analysis EFH-543 2.1 EFH-129 1.8 EFH-544 6.5 EFH-130 2.9 EFH-545 5.0 EFH-131 6.0 EFH-545 2.9

- u a -

~

. o .

(

EW PIPING /_ SUPPORT REANALYSIS

SUMMARY

1 cont'd.) l l

OUTSIDE CONTAI'.tMENT 1 cont'd. J, Systems may be classed as operable on an interim basis if the factors of safety compared to ultimate strengths is less than the original design but equal to or greater than two. ( Ref.

NRC Bulletin 79-02 Supplement No. 1 ). Only two supports

( EFH-129 and EFH-132 ) exhibited safety factors less than two during the elevated temperatures.

Actual stresses on pipe support components ( basepistes )

for EFH-129 and EFH-130 exceeded OBE stress allowable of 12,600 psi per ANSI B31.1 1967 Edition .

of 26,900 psi occurred on support EFH-129 .Maximum OBE stress Maximum OBE stress of 18,100 psi occurred on support EFH-130. All other pipe supports effected by thermal transient maintained component ( baseplate ) OBE stresses within the limits of ANSI B31.1 1967 Edition which are very conservative compared to the stress allowables of 27,000 psi referenced in the AISC '

Code through direction of ASME Section III Subsection NF.

Maximum SSE stress of 33,100 psi occurred on support EFH-129.

Although the stress is greater than the allowable 0.9 Sy d (Sy = 36KSI) the stress would have been below yield.

Impairment of system function or pipe rupture not achieved (

1.e. yield of piping not reached. ) All other supports [

y experienced a maximum SSE stress of 20,600 psi.

I L

. .m o OVERVIEW OF PIPING ANALYSIS EFFORT

( outside Containment )

Piping analysis packages CR-44 and CR-46A were combined

( modeled through anchor restraint EFH-132 ) for the deadweight and thermal analyses. The seismic analysis was defined at the separation points per the original Piping Analysis packages CR-44 and CR-46A.

The degraded pipe support ancnor EFH-126 was modoled as a three directional restraint ( forces Fx, Fz, and moment My )

due to the shear capacity of the anchor bolts existing in the degraded condition. In addition, rigid rod support EFH-128 was removed from the piping analysis due to it's inability to resist the resulting compressive loads generated at this point.

To approximate the thermal transition that occurs within the -

piping, the following temperatures were modeled into the ,

thermal analysis: 2-480 Degrees F - From Pen. 4109 to support EFH-126 to valve FWV-34, and from the tee at analysis node 24 to the elbow above support EFH-5A3, 375 Degrees F - From the tee below EFH-126 to support EFH-132 continuing to and including the flange located east of EFH-133.

255 Degrees F - From the flange located east of EFH-133 to valve EFV-18.

150 Degrees F - From the elbow above support EFH-543 to support EFH-532 and from ETV-18 to support EFH-83.

70 Degrees F - All other affected piping outside containment.

Piping Analysis Results ( Stress Sunsaary )

Primary Stresses Maximum Deadweight and Pressure Stress s 7271 psi 9 Node 1181 Maximum Seismic ( SSE ) Stress = 9174 psi 9 Node 190 16,445 psi Secondary Stresses Maximum Thermal Stress = 14,427 psi 0 Node 1183 A

l 1

EMERGENCY FEEDWATER PENETRATION ANALYSIS

SUMMARY

Initial Design Basis l l

The containment penetration for the emergency feedwater piping consists of a 14-inch sleeve, 2-inch thick containment l side closure plate, 3/4 inch exterior closure plate and a 6 )

inch schedule 160 process pipe. Containment boundary is l formed by the 2-inch closure plate and full penetration welds at the process pipe and penetration sleeve interface to the l

closure plate. The exterior closure plate was provided to facilitate initial penetration functional testing at construction.

Initial design of the penetration assembly and anchorage will  ;

withstand the developed full plastic moment capacity of the '

l process pipe. Axial and shear loads were taken as a force equal to the system pressure times the full flow area of the -

process pipe. For penetration No. 109 these values are: -

l Mu = 1,002,000 in-lbs Vu = 32,600 lbs Pu = 32,600 lbs Tu = 1,090,000 in-lbs Temperature Transien.t Evaluations ( End Plates )

Effects of the thermal transient on the penetration were i evaluated by considering an effective differential l temperature between the containment sleeve and process pipe of 300 degrees Fahrenheit. The actual differential temperature was measured at 208 degrees Fahrenheit during the thermal transient event.

The closure plate flexibility was calculated to define the axial restraint applied to the process pipe within the boundary of the penetration assembly. Initial elastic analysis results indicate development of an axial load of 130,700 lbs. Corresponding stresses within the 2-inch closure plate were determined to be within code limits. The resultant stresses in the 3/4 inch closure plate indicated that elastic capacity was exceeded and plastic response of the 3/4 inch ,

closure plate would be experienced. Axial forces developed in the process pipe when plastic response of the closure plate starts is approximately 28,000 lbs. The corresponding stress in the 2 inch closure plate at this load is 6,000 psi.

External piping system loads induce an additional stress of 5,000 psi in the closure plate. Therefore, a total stress of 11,000 psi was therefore induced on the 2 inch closure plate.

t n @lr

, ,o .

l OVERALL PENETRATION ASSEMBLY PERFORMANCE With the exception of the exterior closure plate on the Type III ( cold ) penetration, there are no physical differences between the Type II ( hot ) penetration and the Type III penetration. The penetration sleeve details, concrete anchorage methods, and the containment liner penetration joint arrangement are identical for the two classes of penetrations. Maximum process pipe temperatures for the Type 1 II penetrations are 600 degrees Fahrenheit which are higher than measured therma] transients for the emergency feedwater system ( measured cc approximately 195 degrees Fahrenheit ) .

Based upon these similarities, the Type III Penetration Nc. 109 has not been subjected to any loading beyond the original design basis.

A detailed steady-state thermal analysis of the containment -

1 structure to penetration interface was performed. Based upon a process pipe temperature of 480 degrees Fahrenheit, may.' Am '-

concrete temperatures are 214 degrees Fahrenheit. The

• concrete temperatures were below 200 degrees Fahrenheit within 3 inches of the penetration sleeve to containment structure interface. These concrete temperatures are in agreement with Appendix A of ACI 349/359 for localized concrete temperatures adjacent to penetration areas which allow for 200 degrees Fahrenheit for long term operation and 350 degrees Fahrenheit for short term emergency considerations.

1

l . a . -1 i

1 2

1 1

\

' t k'H/c % . *;< . .

. -  : . . << l

. . r . ,, . .

.o.. . .,..

~ . ;. < , . -

• 4

\

.\

I4*4 f '.' '.i. e .' * ', **'

.'v'.,

1 u ...' A " ,* * )

(

• - . .. t.s , ,. . o..

w .

.r.. '

}

3,g ,,M

,g"q h

v/ y ,, ," ^ 8 v/ " o a g ,,,,

4

' ,q ,i,< y ,

I I! !' d I

lf I'I tli l

ll ll lI il I

I ll lll l i l ll eg t il l l lI

' l' ^i il l i I

ga '

' I ,

l -

__ l (Scy go) ,._ / ".. . " ,

// /; , U 3,, ,, u. .a u fg

• O lMERMgotag . . ~,. .1 j' *- l * - , _

S REACTOp auitomo .

*? ;,. ' s '.

p,#

t I

.' e .-

\

g \_

T 20 -

__ _ . _ . . _ _ _ , - _ _.