3F0788-12, Forwards Suppl to Util Re Emergency Feedwater Leaks & Summary of Presentation During 880708 Meeting in Rockville,Md,Per 880706 Confirmation of Action Ltr. Corrective Actions Associated W/Stated Insp Rept

Forwards Suppl to Util Re Emergency Feedwater Leaks & Summary of Presentation During 880708 Meeting in Rockville,Md,Per 880706 Confirmation of Action Ltr. Corrective Actions Associated W/Stated Insp Rept

ML20151D791
Person / Time
Site:	Crystal River
Issue date:	07/14/1988
From:	Ken Wilson FLORIDA POWER CORP.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
3F0788-12, 3F788-12, CAL, NUDOCS 8807250237
Download: ML20151D791 (21)

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ees Power C ORP O A AtIO N July 14, 1988 3F0788-12 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 submitted on June 28, 1988, concerning the emergency feedwater leaks and summarizes the 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 discussed further at a follow-up meeting yet to be scheduled. The issues associated with our corrective actions for previous events were discussed at the July 8 meeting and will be the subject of separate correspondence with Region II associated with Inspection Report 88-18.

If you have any questions, please contact this office.

Sincerely, n

eSo Wh%

wo1 K.R. Wilson, Manager S8 Nuclear Licensing KRW/REF/dhd Attachment xc: Regional Administrator, Region II O 8

Senior Resident Inspector { g

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GENT:RAL OFFICE: 3201 Thirty-fourth Street South

• P.O. Box 14042

• St. Petersburg, Florida 33733 * (813) 866-5151 A Florida Progress Company

. a ENFIRRTIN OF ACTIN IEPITR RM4ARY OF JUIX 8,1988 IHESENIRTim Backrzrnurd m Curret EFW Line Heat tm Event:

During Refuel VI, several metgency feutsr (EFW) systm valves underwent refurbishing. 7hese efforts included the repair of EEV-33 aM the inspection of EFV's 15,16,17, aM 18 (see Attachment 1 flow schmatic) . 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 uns m m ful. The unit was returned to service on January 10, 1988.

Crystal River Unit 3 experienced a reactor trip on February 28, 1988, due to a malfunction in the feedwater system which resulted in an EFW injection (IER 88-

06) . The plant entered a ! ODE 3 outage for 4 days to investigate the cause of the trip. The unit was returned to service on March 3,1988, but limited to 60% power. The unit was removed frm service on March 7 aM entered LODE 2 to effect repairs on the main feedwater flow control valve. Durirg the startup on March 8, the Operations staff discovered all fcur EFW lines hot. The check valves in the EFW lines wers apparently allowiry scxne backflow causing the heat up of the EFW lines. Estimates of the torperatures and conditions of the lines were provided by Operations to Engineerirg. Follow-up inspections of the EFW lines determined that the tecperature had returned to nomal following a cool down process khich involved flcuing a small arount of emergency feedwater through the lines. This action evidently caused the leakiry check valves in the lines to rescat since these lines remained cool until May.

On May 27, 1988, during a Itutine plant walk down, the Shift Operations 1bchnical Advisor (SorA) discovered the eneraancy feedwater pipirg near flow control valve EFV-55 to be warmer than normal, indicatiry check valve FWV-43 was not preventirg backflow into the emergency foodwater line frua the steam generator. This cordition was dimwa with the Shift Supervisor on duty ard increased awareness of the cordition was maintained by site personnel. By June 19, the ter:perature of the line frm the reactor buildify penetration to EFV-18 was hot enough to cause scrre flashing of the water leaki q frm EFV-18. A lion-Confoming Operations Report (lic)R 88-81) was written on June 19, ard upon further investigation a one hour report was initiated on June 21. The report was cased on exceeding the design basis tenperature for the emergency feedwater line. Prior to this time, the line was thought to have been designed and analyzed for nuch higher terperatures.

As a result of this heat up, visual inspections ard enginocring analyses sure conducted. These activities irduded:

o Visual inspections of the penetration ard walkdowns of a-ible pipirq ard supports by Engineerirg personnel o Pressure test of Penetration $109 o Dye penetrant inspection of high stressed kuld on Penetration #109 i

o Themal stress analyals on penetration steel 1

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o Review of temperature effects on penetration cuh Le o 'Ihermal analyses of EFW piping inside and outside of reactor building o Installation of a tarporary m:xilfication to reinject leakage fran EEV-18 to stop the EFW line heat up.

o Reevaluation of stranaam caused by seismic, daadweight arxl internal praa=we forces as a result of the high temperature condition.

o Radiograph of check valve FW-43 o Repair of pipe restraint FE -126 o Leak repair of EEV-18 Details of these activities were provided during teleconferences and in a letter to the NRC dated June 28, 1988. Additional information concerning the penetration analyses, the thermal flexibility analyses of the piping, the pipe support evaluations, and the leak repair of EFV-18 is provided below.

Penetratim Analyses:

'1he penetration consists of a 14" sleeve and a 2" thick end plate inside the reactor building. A 6" Schedule 160 prmana pipe pa=== through the center of the sleeve arxl end plate. Welded lugs secure the sleeve to the concrete containment wall. A 3/4" plate is welded on the sleeve to provide a closure plate to facilitate testing of the penetration. 'Ihe 3/4" plate is not considered part of the containment bourxlary, and is not credited as a pipe support in the piping analysis.

A two dimensional, steady state heat transfer model was developed to determine the maxinaan anticipated concrete tenperature attained near the penetration inside the Reactor Building (RB) . Raaai on a maximan prmaaa pipe temperature

. (measured) of 480' F, a mavi== concrete temperature of 214'F was predicted.

'Ibe model was validated by ocmparing its predications with measured values obtained at a location above the penetration outside the RB. Excellent agreenent was found (mavi== measured was 195'F) .

Although the overall concrete design temperature for G-3 is 150'F(based on ACI i 318-1963), FPC considers the criteria established in ACI 349 & 359 to be relevant for this transient situation. ACI 349/359, A@endix A, permits localized concrete tatperatures adjacent to penetrations to be 200*F for long j term periods (continuous) and 350*F for short term emergency situations.

'Iberefore, FPC concluded and visual inspection confimed the concrete was not adversely affected by this elevated temperature ocn11 tion.

'Ihe penetration sleeve and erxl plates were analyzed by conservatively ===irg a 300*F di*ferential tenperature between the sleeve and prma== pipe. 'Ihis aW the actual measured differential tenperature of 208'F. Under a 300*F differeritial temperature, it was determined that the 3/4" closure plate would be stressed beyond its elastic limit (28,000 pounds-force) . 'Ihe corresporvlirg stress in the end plate frca this load is 6,000 psi. 'Ihis stress, when added to the stress frun the piping loads of 5,000 psi, are well within the allcwable stress of 45,000 psi, thereby assuring that contalment integrity was never cx:mprunised.

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W 3/4" closure plate was 1caded beyond its elastic range aM was performing in the plastic range, however, FPC does not consider this to be a problan since its only function is to facilitate testing of the penetration. It has not been taken credit for in maintaining containment integrity or as a pipe support.

An additional inspection and dye penetrant test were conductM to demonstrate no degradation has occurr M to the 3/4" plate.

Pinina Analysis:

A thermal flexibility analysis was performed on the piping inside and outside of the Reactor Building (RB) . 'Ihe analysis was perfIrmed at the elevated temperatures.

'Ihe piping inside the RB was analyzed at a taperature of (80*F from Ibnetration #109 to a point midway to the steam generator, and at 590'F for the remaining portion. 'Ihis temperature profile is a conservative assu:rption since the actual gradient between the steam generator aM the penetration would be more gentle. 'Ihis portion of tha stress analysis demonstrated that the pipirq inside the RB frun Penetration #109 to the steam generator was not stressed above ANSI B31.1,1967 code allowables during the elevated taperature corxlition. Maxinzm stress in the piping was calculated to be 17,256 psi shich is below the allowable of 22,500 psi.

A thermal analysis was performed on the pipiry outside the RB at the elevated ta peratures measured from EFV-18 to Penetration #109. 'Ihe system supports were modeled to reflect the as fouM coMition based on the viscal inspections.

'Ibe results of the analysis inlicated that the piping outside the RB was not subjected to stresses in excess of code allowables. W ma::1 mum calculated thermal pipe stress was 14,427 psi which was less than the 22,500 psi allowable.

A pressure, deadweight, aM seismic (SSE) load analysis of the piping outside the RB was performtri with EFH-126 rodeled in the damaged ocniition aM providing only translational ard torsional restraint. Under these load canbimtions, the maximum calculated pipe stress was 16,445 psi khich was less than the allcwable 18,000 psi.

Eipe Stzmort Evaluations:

'Jhe effects of the loads irposed by the pipirg on the pipe supports were evaluated. 'Ibe pipe supports inside the reactor tuilding here not subjected to stresses in excess of B31.1 ocde allowables durirq the elevatcd tercrature ccadition. In acklition, only one (1) pipe support on the subject piping inside i the RB utilized anchor bolts. UMer the worst case canbined loading, the anchor bolt safety factor was reduccd to 5.6 ccrpared to the mininun 4.0 cpecified in the NRC Bulletin 79-02. 'Iherefore, pipe supports inside the RB remained fully functicml during the elevated tcrperature ocniition.

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'Ibe supports on the pipirg outside the RB were also evaluated. Base plate i

str- in two supports (EFH-129,130) wtuld have exceeded B31.1 allowables i durirg seismic events. 'Ihc worst case value of 33,100 was below the yield stren]th of 36,000 psi during this postulated loadiry condition.

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e During the postulated SSE loading coMition, anchor bolt safety factors for two supports would have been below the minimum of 2.0 as specified in Bulletin 79-02 . We mininum calculated safety factor was 1.7 at DH-132, which is above 1.0 and does not constitute a failure of the anchorage. It is inportant to eriphasize that these are worst case load ocanbinations based on seismic, deadweight and thental pipe loads. Actual loading during the elevated t p ture conditions, excluding seismic, was within allowables at all times.

D V-18 Isak Rnonir:

On June 28, 1988, a testporary nodification was made on DV-18 to seal the body to bonnet leak. A low testperature, flexible sealant was injected by means of a manually operated pressure gun into the spaces on top of the bonnet and inside the lock ring of EEV-18 which totally eliminated the leak. A diagram of EFV-18 is provided in Attachment 2. % e sealant was injected into the valve first tnxugh the modified cap bolts to seal the interior space aM bolt screw holes io ea "A"), and ww1d through the two valve vent ports to seal the spaces in cN valve lock ring and cap (area "B"). W e scalant was injected at pressures talow the system pressure to preclude sealant from entering tne EEW line except for the final few strokes of the gun. %ese final strokes nust me at an injection pressure exceeding system pressure in order to cortpress the sealant aM stop the leak. Bis method minimizes the potential for sealant to enter the valve internals. M EEW system flow test of approximately 50 gpm was used to confim valve funct ; M unaffected by the use of the sealant. A test of the valve integrity wa4 C W acted by placing the EFW system into recirculation thereby, generating max 1___. operating pressure. %e chemistry of the sealant was certified to be within acceptable values and ccrpatible with valve material.

Previous Exoerience with DW Line Heat 13:

A review of previous DW events revealed three (3) License Event Reports (IER's) aM two (2) Nonconforming Operations Reports (NOOR) had been written hwnting the problem with the flow transducers on the EFW lines due to high tenperatures. During the time frame of these events, IW-43 was suspected of backleakage which resulted in the EEW line beat up. % e flow paths for the leakage were not identified in the documents. A sumary of previous events is provided in Attachment 3.

Planned inspections of DW check valves were perfomed during o.Itages in 1980, 1985 and 1987. We 1985 aM 1987 inspections were planned IN's instituted as a result of SOER 84-3. A review of the maintenance history of the DW system valves was perfonmd and EW-43 and DV-33 were identified as having a trend of poor perfomance frun the standpoint of backleakage and leaks to atrosphere.

Evaluations have been perfomed in acconlance with the Institute of Nuclear Power Operations (INFO), Significant Operating Event Repcrt (SOER) 86-3, which also identified these valves as candidates for additicn4 inspcction and testing.

An announced NRC inspection was conducted at CR-3 durirg Octooer 5-9 1987, to detemine the current status of check valve test programs as a subset of the industry as a whole and to detemine the responsiveness of the industry to the DUO reocamendations for inproving check valve testing programs (SOER 86-3) .

During the inspection, the NRC identified DV-18 and EW-43 as having mintemnce probles and as being inchrled in the SOER response program.

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Future Accim:

Florida Ebwer Corporation on, or befom, October 15, 1988, will pennanently repair EFV-18 and EFV-33 body to bonnet leaks, and will inspect and repair, as necessary, FW-43 and EFV-16 seat leakage. In acklition, FW-44 will be inspected and repaired, as ne ssary, based on the results of the inspions and repair efforts of FW-43. F N is currently evaluating the a @ ropriate post ,

maintenance test and acceptance criteria for FW-43 and FW-44. j l

Hourly tertperature readirgs of the EFW pro ss pipe through Penetration #109 j will be maintained until a permanent fix is made. 'Ihe once per shift check of l EFW purips discharge piping will continue, as well as refueling interval l inspections of EFW check valves. A ccrtplete evaluation of the maintenance history on these valves will be perfonned and utilized in conjunction with the ,

INFO SOER 86-3 evaluations to establish reccanendations for possible changes to I preventative maintenance prwrars. '

A review will be perforwxi to upgrade the EFW piping for actual terrperatures eypected in the system. 'Ihis review will include high energy line, stratification and water hamer effects as appropriate.

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CIMFII59&ICH OF ACI'ICH IRL'IER RESIDEB Besomses to Soecific Confirmation of Action Ietter Itms:

Itan:

A discussion of the effect of potential EFW system degradation on the reliability of the systan aM on larg-term plant safety.

Nepmse:

The primary adverse consequence was EFW reliability. Florida Power Corporation maintains that the EFW reliability was unaffected since the piping, penetration and supports were demonstrated through analyses to be capable of performing their inteMed functions within the design allowables excluding seismic loads.

The piping has been returned to within the design tanperature and will be maintained within this specification, thereby, alleviating the additional thermal stresses experienced during the abnonnal heat up events. The valves and system performance were and continue to be unaffected by the recirculation system or the injection sealant as discussed in earlier correspondence and the presentation sumary contained herein.

Itan:

A description of the repair of EFV-33 and of the injected sealant repair of EFV-18, results of the repairs, and description and rest 11ts of testing to verify cmponent/ system operability after repairs.

• somse:

The on-line repair of EEV-33 is under evaluation. Isakage frm EEV-33 has decreased and will continue to be monitored. The injected sealant repair of EFV-13 and subsequent results and tests are discussed in detsil in the previous pages as part of the July 8, presentation summary.

Itan:

The results of all thermal and stress analyses of all involved pipirg, penetrations, and supports.

Respmse:

The results of the thennal and stress analyses of the penetration, piping and supports are discussed in the July 8, presentation summary. Additional analytical information is provided in AttacA u t 4.

Itan:

Evaluation of the safety implications of the valve leakage and EFW system overte@erature and of the EFW system with the interim fix and the temporary repairs in place.

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nesponse:

W e EFV system valve leakage did not pose any safety concerns. We leak rate did not eyr,aM 0.9 gpn and had no adverse effects on the EFW flw to the steam generators. %e potential for water hanmer was considered unaffected by these events due to the follwing:

o % e piping diameter was small o he piping has short horizontal runs and numerous beds o he piping has rigid supports at the majority of the pipe bends o % e EFW enters into the steam space in the steam generator o %e majority of the piping remained filled with water because the tertrarature was belw the saturation temperature of 540*F for a '

majority of the piping.

o %e piping generally rises frm the EFW pumps to the steam generator.

We potential for steam biniing f the EFW pmps was also considered.

Monitoring was initially conducted o a once per shift basis and later upgraded to hourly to quickly identify coMitions which could lead to steam binding. Eere was never a problem or threat of steam binding of the EFW punps as a result of the leaks since other valves in the lines were preventing back f1w and heat up of the system. We FW-43 back-leakage was not considered to be a significant safety con rn sin the valve was mnsidered operable for EFW injection aM minor backleakage was of little consequence as long as the EFW piping and pumps remained cool.

W e m ntainment integrity issue was not considered significant since the piping and steam generator provided system integrity in the RB (closed system inside containment). For a steam generator tube rupture event, the off site dose consequences remain well within design basis for this accident.

l %e tempcrary injection system was carefully designed to consider EFW system isolation, seismic restraint and relief valve needs in order to adequately protect the reliability of the EFW system. We tenporary EFW-18 leak repair was discussed in the July 8 presentation summary.

Itesn:

l Verification that EFW cmponents have not been damaged by overtearperature events and can reliably perform their safety functions, nesponse:

he only adverse consequences not addressed elsewhere involved EEV-55 ard EFV-

! 57. W e effects of higher than expected temperatures on the qualification life l of EFV-55 and EFV-57 were evaluated, and sufficient margin is retained for continued operation. Early repair / replacement will be evaluated as part of our l

normal EQ maintenance program.

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Itan:

A dimicaion of the possible need for a more ocmprehensive inspection ard repair program based on an evaluation of current and past valve leakges.

E==ipus =:= :

Florida Power Corporation is continuing to evaluate and develop a more ccmprehensive inspection and repair program for check valves. FPC will factor the EFW events into that program. These events do Dg_t contradict the priorities identified by that program (e.g. these were all high priority valves for enhanced efforts). FPC's detailed response to previous check valve leakage events will be discussed more throughly in separate correspondence.

1 8

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ATTACHMENT 3 s SUPMARY OF PREVIOUS EXPERIENCE l

l DME PROBLEM CAUSE LER's l

1/19/83 82-076 HEAT FAILURE OF FLOW TRANSDUCER FW-43 LEAK

8/19/83 83-029 FAILURE OF FLOW TRANSDUCER HIGH HEAT FAILURE OF SEALANT l MOUNTING

8/24/84 83-043 FAILURE OF FLOW TRANSDUCER FW-43 LEAK 7 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) 10/15/82 82-25 17 MIN FEED TO HIGH N0ZZLE POST TRIP FAILURE OF FW-39 FOLLOWING RESTART OF EFP D

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5 ATTACHMEN1' 4 DETAILED ANALYSES i

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i EMERGENCY FEEDWATER PIPING f SUPPORT REANALYSIS

SUMMARY

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 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 Factor for EFH-025A ( only support attached to concrete in the effected piping ) under deadweight, pressure, and SSE loading conditions = 19 .

- During Elevated Temperatures All actual stresses 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 .

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OVERVIEW OF PIPING ANALYSIS

( Inside Containment )

6" Diameter Emergency Feedwater Piping from Penetration No.

109 to Steam Generator 3D was analyzed for thermal effects.

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 point 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 acceptable from an allowable stress of 22,500 psi provided in ANSI B31.1.0 1967 Edition and B&W Dw'g. No. 108031, Rev. O.

Maximum Primary Stress

- Maximum Deadweight & Pressure Stress = 5,665 psi @ node 32 Maximum Seismic ( SSE ) = 11,700 psi @ 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.

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EMERGENCY FEEDWATER PIPING /_ SUPPORT REANALYSIS

SUMMARY

OUTSIDE 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 = 7,906 psi.

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 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 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 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 = 14, 427 psi which is less than the allowable stress of 22,500 psi permitted in ANSI 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 pai 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 Support Safety Factor EFH-125 5.5 EFH-133 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-546 2.9 EFH-132 1.7 EFW PIPING / }UPPORT REANALYSIS

SUMMARY

1 cont'd.)

OUTSIDE CONTAINMENT 1 cont'd. 1 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 ( baseplates )

for EFH-129 and EFH-130 exceeded OBE stress allowable of 12,600 psi per ANSI B31.1 1967 Edition . Maximum OBE stress of 26,900 psi occurred on support EFH-129 . Maximum OBE stress of 18,100 psi occurred on support EFH-130. All other pipe supports effacted by thermal transient maintained componcat ( baseplate ) OBE stresses within the liinits 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 (Sy = 36KSI) the stress would have been below yield.

Impairment of system function or pipe rupturo not achieved (

i.e. yield of piping not reached. ) All other supports experienced a maximum SSE stress of 20,600 psi.

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OVERVIEW OF PIPING AitEY:}IS EFFORT

( Outside Containzant )

Piping analysis packages CR-44 and CR-36A were combined

( codeled through anchor restraint EFH-132 ) for the deadweight and therma.1 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 anchor EFH-126 was modeled 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:

480 Degrees F - From Pen. #109 to support EFH-126 to valve FWV-34, and from the tee at analysis node 24 to the elbow above support EFH-543.

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 Degraes F - From the elbow above support EFH-543 to support EFH-532 and from EFV-18 to support EFH-83.

70 Degrees F - All other affected piping outside containment.

Piping Analysis Results ( Stress Summary )

Primary Stresses Maximum Deadweight and Pressure Stress = 7271 psi @ Node 1181 Maximum Seismic ( SSE ) Stress = 9174 psi @ Node 190 16,445 psi Secondary Stresses Maximum Thermal Stress = 14,427 psi @ Node 1183 17 -

O EMERGENCY FEEDWATER PENETRATION ANALYSIS

SUMMARY

Initial Design Basis The containment penetration for the emergency feedwater piping consists of a 14-inch sleeve, 2-inch thick containment side closure plate, 3/4 inch exterior closure plate and a 6 inch schedule 160 process pipe. Containment boundary is formed by the 2-inch closure plate and full penetration welds at the process pipe and penetration sleeve interface to the closure plate. The exterior closure plate was provided to facilitate initial penetration functional testing at construction.

Initial design v. the penetration assembly and anchorage will withstand the de veloped full plastic moment capacity of the 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:

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

Effects of the thermal transient on the penetration were evaluated by considering an effective differential 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.

l The closure plate flexibility was calculated to define the I axial restraint applied to the process pipe within the l 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 l capacity was exceeded and plastic response of the 3/4 inch t

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.

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 II penetrations are 600 degrees Fahrenheit which are higher than measured thermal transients for the emergency feedwater system ( measured at approximately 195 degrees Fr.hrenheit ) .

Based upon these similarities, the Type III Penetration No. 109 has not been subjected to any loading beyond the original design basis.

A detailed steady-state thermal analysis of the containment structure to penetration interface was performed. Based upon a process pipe temperature of 480 degrees Fahrenheit, maximum 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.

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