ML20137M266

Rev 0 to JPN-PSL-SENP-94-079, Engineering Evaluation, Assessment of ECCS Suction Piping Crossite Due to Design of Naoh Spray Additive Sys

ML20137M266
Person / Time
Site:	Saint Lucie
Issue date:	11/30/1994
From:	FLORIDA POWER & LIGHT CO.
To:
Shared Package
ML20137M095	List: ML20137M089 ML20137M161 ML20137M177 ML20137M204 ML20137M212 ML20137M224 ML20137M250 ML20137M266 ML20137M358 ML20137M416 ML20137M495 ML20137M532 ML20137M541 ML20137M711 ML20137M763 ML20137M803 ML20137M907 ML20137P297 ML20137P306 ML20137P338 ML20137P343 ML20137P355 ML20137P365 ML20137P417
References
FOIA-96-485 JPN-PSL-SENP-94, JPN-PSL-SENP-94-079, JPN-PSL-SENP-94-79, NUDOCS 9704080044
Download: ML20137M266 (28)

Text

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l FLORIDA POWER & LIGET Co  !

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ENGINEERING EVALUATION l

A8sESSMENT OF BCC8 SUCTION PIPING CROSSTIE DUE TO DESIGN OF NaON 87 RAY ADDITIVE SYSTEN I

ST LUCZE NUCLEAR PLANT l

UNIT 1 1

JPM-PSL-8ENP=94-079 i

i REVISION 0 BAPETY RELATED .

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' JPN-PEL-5ENP-94-079 REVISION O .

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REVIEW AND APPROVAL RECORD t

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MaOH SPRAY ADDITIVE SYsits LICENSING LEAD DISCIPLINC i

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! TABLE OF CONTENTS l

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SECTION M 2AQZ

-- Cover 1 ,

Review and Approval Record 2 l 4 Table of Contents 3 i l

1.0 Purpose;and Description 4 I l ~

2.0 Background 4

'3. 0 Design Bases 5 .

4.0 Analyses of the Event 11 5.0 Conclus' ions

• 14 I

6.0 Verification Summary 14 7.0 References 15 Attachments None i ,

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JPN-PSL-SENP-94-079

• l REVISION 0 PAGE 4 OF 15 s

! 1.O PURPOSE AMD DESCRIPTION I The purpose'of this evaluation is to assess the significance to plant operation and safety associated with emergency core cooling L system (ECCS) suction piping crosstie due to theOndesign of the October 20,

Sodium Hydroxide (NaOH) spray additive system.

1994, Unit.1.was in Mode 1 and operating at 100% power.

' Differential, pressure testing of a motor operated valve resulto'd

-.in the lif ting of a safety relief valve on-the suction of the

! ECCS supply header. On October 23, 1994, an Initial Assessment of operability (Ref 3) determined that this safety relief valve could lift under certain accident conditions-and result in sump inventory loss to the reactor auxiliary building (RAB) in excess of ECCS design external system leakage.

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. 2.0 BADEGROUND

! Examination of the alignment used to perform the motor operated

' valve test revealed a flow path from the discharge of the 1B containment spray pump to the suction of the 1A containment spray-j pump through.a common header in the NaOH spray additive system.-

With the IB containment spray pump operating and all of the A ECCS pumps secured, the A train ECCS suction piping, common to the A train ECCS pumps, was pressurized through a spray additive system common line. As part of the investigation of the event, the 1B low pressure safety injection pump was aligned to the B trein containment spray header and discharged through the B 4

shutdown cooling heat exchanger with the same result, lifting of the A -train safety relief I-SR-07-1A (60 ps'ig lift setpoint) .

The maximum pressure observed at the 1A low pressure safety

injection discharge header was 80 psig. This event was documented in St. Lucie Action Request (STAR) 1-94100259 (Ref 4).

The initial assessment determined that the components affected by the pressurization are capable of withstanding considerably higher pressures (Ref 3). As such, the suction piping and

components were not adversely affected as a result of the event.

L However, with regard to the issue of system design, it was concluded that a scenario exists which may affect the results of a design basis ev' int. The existing design could result in an ECCS suctioniheader relief valve lifting for the event consisting

' of a LOCA (SIAS and CSAS) with a loss of off-site power (LOOP) and one emergency diesel generator (EDG) failing to operate.

This relief valve could potentially release containment sump

inventory following recirculation actuation (RAS) which is outside design basis event bounding conditions for ECCS external system leakage (Ref 2, Section 15.4.1.7, and 15.4.1.8). This scenario could result in a condition outside the accepted design leakage flowLlimits for the engineered safeguards systems (Ref 2, Table 15.4.1-2, Section 15.4.1.7). The system design conditien i has existed since the NaOH spray additive system was backfit to

. Unit 1 in 1978 (Ref 9). Based on the initial assessment, plant 4 ' management made a one hour non-emergency notification to the Nuclear Regulatory Commission in accordance with 10 CFR 50.72.

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i jL l PAGE 5 0F 15 i L Following the, initial assessment, a-safety evaluation (Ref 5) j- determined acceptability of higher pressures in the ECCS suction l piping in order to disable relief valves I-SR-07-1A and I-SR  !

1B for the remaining Unit a cycle (approximately 1 week). This i interim measure was implemented to preclude the possibility of l excessive ECCS leakage. The NaoM spray additive system eductor j

piping was modified during-the 1994 refueling The safetycutage relieftovalves eliminate i

the cross connect described herein.

j- have been returned to service.  :

t l A design review for. train cross connections was performed for

~ECCS (Ref 7).. Except for the common line identified in the NaoM i

i spray additive system, the review identified no other train cross '

connections which do not meet single failure criteria as applied to St. Lucie Unit 1 & 2, as discussed La the respective FSARs and * ,

current versions of the design basis divuments (DBDs).

! The architect engineer, Raytheon Engineers & Constructorswas notified (Ebasco),  ;

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4 deficiency by a letter (Ref 16). This letter requested that Raythgonreviewthedesignfor10CFRPart21 significance.

A Licansee Event Report (LER) was submitted describing the e: vent, As

! the root cause,.and some of the preliminary analysis (Ref 65 i- part of ths LER corrective actions, an assessment of the safety j This consequences,and implications of this design is required.

engineering evaluation is a basis for that assessment and will be l

' used as a source document for a supplemental LER.

I This engineering evaluation involves engineered safoguards l-systems and is therefore classified as safety related.

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i 3.0 DESIGN BASEE ' REVIEW l-j ECCE Deslan Bases (Raf 2. Baction 6.3.1.11 i

The emergency core cooling system is designed to provide core cooling in the unlikely Theevent following of a LoCA designfor all postulated criteria form the basis pipe breaks in the RCS.

for design of the ECCS I

a)

The safety functions defined in TSAR Section 6.3.1.1 must be

', accomplished assuming the failure of a single active component during the injection mode of operation or the single failure of an active or passivo component during the l

' recirculation mode of operatic.n.

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' b) The design of the safety injection system must provide for inspection and testing of couponents and subsystems to l

ensure their availability and reliable operation. i I

i c) All components of the. safety injection system and associated

' critical instrumentation which must operate following a LoCA

' are designed tw operate in the environment to which they would be exposed in the event of 'a Lob..

eena OcHDC 9T. LUCIE  !!/30/94 16: 40 P.  !

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' JPN-PSL-SENP-94-073 REVICZGN o' PAGE-6 or 15 ' a!

-t d) All components of the system are designed as seismic class I l to withstand the forces of the design basis earthquake (DSE). i

-The SER'(Ref_10, Section 6.3) evaluated the design of the.ECCS j

and determined that."The ECCS subsystems provided are of such j

i number, diversity, reliability, and redundancy The single. failure that no sing s inadequate cooling of the reactor core".Section 6.3.3.2 and Table 6.3-analysis 3.

is described in the FSAR,In addition, ECCS design required that performance conform to criteria set forth in subparagraph (b) of

[ Section 50.46, 10CFR Part 50, Acceptance' Criteria for ECCS Supplement 2 t4 for the Light Water Cooled Nuclear Power Reactors.

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SER (Ref 10,,Section 6.3) concluded acceptability of the 2005

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performance, and therefore accept 4bility of the ECCS.,

!- ,1 Eccs Suetion Handar Damian i j

-The.ECCS suction piping is designed in accordance with The code the not does Class II.

l requirements of USAS B31.7-1969, provide a specific requirement

' portion of the system. The relief valves.of However, concern are valves the relief not ,

specifically.. addressed in the FSAR. l were installed to address a concern over the interface between i low pressure suction piping and the portion of suction piping '

' used for shutdown cooling. This protected against leakage across f motor operated valves and/or the failure to isolate this portion of the system prior to initiating shutdown cooling (Ref 5).

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l-ECCS External Svatam Laakaae l

l i: External system leakage for ECCS components is descri i- L.P. It discusses ECCS external system leakage Document (DBD-SDC-1).

due to a passive failure during the recirculation pha'se following

' a LOCA as limited to leakage through a failed pump seml or valve

! packing. Gross failure of ECCS piping is not considered credible.

l FSAR Section 15.4.1.7 and Table 15.4.1-2 defines leakage from engineered safety features (EST) components outside containment.

The maximumileakage to the ECCS area is defined as 2 liters per hour.

All ESF components *containing recirculating sump water are i within the controlled ventilation area served by the ECCS area ventilationisystem (Ref 2, Section 9.4.3).

I L The design flow of the spray additive system eductor is 128 gp ,

(Refapproximately are 5). equal with a relief valve differential pressure of 30 psid.' This is consistent with the pressure of t

.approximately 80 psig observed during investigation of the even

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i using the 1B LPSI pump (Ret 3). .

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1. ROM USNRC 5T. LUCIE 11/30/94 16249 P. 5 F '

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JPN-PSL-SENP-94-079 (8' REVZCICH 0 J-PAGE 7 OF 15 The safety relief valve RAB post-LOCA leakage environment-A review of l- considers recirculation water temperature and dose.

the scenarios and the estimated. dose indicates that equipment

potentially affected by the leakage would operate within thebou Based on the, design of the RAB and the ECCS area ventilation system, the safety related. equipment. located within the.RAB trea potentially affected by the lpakage Although operates.within RAB EQthe bounds of is acceptable, environmental qualification (CQ).

l uncontrolled. leakage if left unchecked could eventually submerge .f safety related equipment, resulting in total loss of ECCS' '

l cooling, and; adversely impact accident mitigation capability.

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i-The RAB floor area which includes the pipe tunnel is provided

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F with plumbing and drainage normally controlled via the wasteAn alterna management system.-

11.2.2.1, returns radioactive leakage from the ECCS pump area to

' the containment' building. T$1s alternate a path is the leakage dedicated system which is

' collection and return system (LCRS),

aligned to the reactor drain tank (RDT) following RAS (Ref 12).

The LCRS requires-instrument air.and non-vital power. This l

i system utilizes the safeguards pump room sump pumps (2 pumps perEach sum to pump back to containment. The sump pumps train)ing operat flow rate of 50.gpm at 100 ft. of head.

i are powered via non-vital MCCs 1A2/182 which would not be available during LOOP scenarios. ,

1 Containment spray discharge header pressures vary depending upon )

scenario, however, NaOH eductorA flow rate rate leakage is relatively of 128 gym is l insensitive to motive pressure.

I therefore assumed in conjunction with The LOOP (nofloor sump pumps l

available)_in evaluating effects. RAB area and pipe i-tunnel flooriarea'can contain the volume of water from a safeti l

relief valve discharging 128 gpm for approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> without adversely affecting the idle ECCS train (train affected by loss of power /EDG failure). The operating train is protected from flooding for approximately is hours assuming wall Leakage l

penetrations. separating A & B trains remain watertight.

if left unchecked could eventually submerge the operating safety

! related equipment, resulting in the total loss of ECCS cooling,

' and adversely impact accident mitigation capability.

! ECCS Arma Ventilatien (Raf 2. Section 9.4.3)

I The ECCS area ventilation system is designed to provide post-LOCA filtration and absorption of fission products in the exhaust air J from areas of the RAB which contain the following equipment: l

, l a) Containment isolation valves; b) ' Low pressure safety injection pumps; c) High pressure _ safety injection pumps; j d) Containment spray pumps; .

i e) Shutdown heat exchangers; f). Piping:which contains recirculating containment sump water l following a LOCA.

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[' FROM USNRC ST. LUCIE 11/30/94 16: 49 P. ?

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JPH-PSL-SENP-94-079- "i REVIDIoM 0 PAGE 8 0F 15 ['

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.The ventilation' system is sized to maintain a slightly negative pressure in the~ engineered safety Uponfeatures arenlwith loss ofenormal respect power, theto surrounding areas of the RAB.

system will be automatically connected to the emergency powerThe fans and da source if required to operate.

t with each of the separate filter trains are powered from separate buses and receive actuation signals from separate SIAS channels.  !

] No single failure will prevent both trains from operating.  !

l The ECC) equipment area is ventilated by passing outside air through it. The ECCS area ventilation system iodine maintains the'ECCS activity equipment area at a. negative pressure, thus, associated with any ECC

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l leakage is passed through the ECCS. system charcoal absorbers.

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i A best estimate dose assessment was performed to evaluate dose '

.i consequences of a LOCA assuming the current design of the sprayEvalua l

additave system (Ref 11).

accident scenarios indicates that no core damage (peak clad l l

i temperature <2,200*F) would result since one train of ECCS l operates to cool the core and ECCS performance is not degraded 1 (Ref 8). As such, the following assumptions were considered for j determination of dose, 100% of the gas gap is released

, a) 100% of.the fuel rods fail,This is conservative since one train of inside containment.

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ECCS operates and meets ECCS performance criteria with only I

limitedifuel clad damage predicted.

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b) '95% of halogens bound as Cs-I, and are not available to become irborne in the ECCS room.

c) The recirculated sump. water temperature is assumed to remain This above 212'F. This allows for a partition value of 10.

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' is conservative since the sump water is recirculated through the shutdown cooling heat exchanger and would remain less l than 212'F (Ret 2, Section 9.4.3.3), at which a partit' ionThis factor of approximately 100 would be more appropriate.

enhanced partition factor reduces offsite dose by a factor of up to 10.

j d) Filter and iodine removal efficiencies are nominally 89%.

This 1s; conservative cince actual removr;l efficiencies are

- tested to 99.9%

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The results of the above assessment indicates thatThe thedoseoffsite to dose is less than 1/10 of 10 CFR part 100 limits.

the exclusion area boundary and the low population zone remain i within a small fraction of 10 CFR Part 100 limits and the dose to the control room would not exceed the conservative values described in the FSAR. However, radiological consequencas could I increase if; flooding left unchecked results in the total loss of

' ', ECCS cooling.

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y. PAGE 9 OF 15 Containment serav Damian Bases (Ref 2, Section 6.2.2.11.

The containment heat removal system consists of the containment spray system and the containment cooling system and is designed h to prevent the containment pressure from exceeding its design value following a LOCA, assuming a single active or passive failure.

I The containment spray system consists of two independent and

- redundant subeystems. The heat removal capacity of either of the two subsystems is adequate to keep the containment pressure and temperature below design values and to-bring the containment pressure below 10 psig within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after any size break in the reactor coolant system piping up to and including a double-

• i ended break of the largest reactor pipe, assuming unobstructed l discharge from.both ends. ~ 1 I

Any of the following combinations of equipment will provide at l least minimum heat removal capability necessary to limit and

reduce the post-accident containment pressure and temperaturet i . .

a) All four containment fan coolers

, b) Either of the two containment spray nubsystems c) One containment spray subsystem in conjunction with two

containment fan coolers l

The containment Spray system has two modes of operation which

" ares

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a) The initial injection mode, during which the system sprays i

borated water from the refueling water storage tank into the containment.

l b) The recirculation mode, which is automatically initiated by the RAS after low level is reached in the refueling water tank. During this mode of operation, suction for the spray pumps is from the containment sump.

I containment spray is automatically initiated by the containment spray actuation signal (CSAS) which is a coincidence of the safety injection actuation signal (sIAs) and the high-high

' containment pressure signal.

' I The SER KRef.10, Section 6.1 & 6.2.2) evaluated the design of engineered safety features (ESF) which includes the containnent spray system. The SER concludes that the containment spray system meets single failure criteria.

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i An evaluation lwas performed to assess the impact.of reduced '

4 containment spray. flow on the transient safety analyses and the containment analyses (Ref 8). Assuming the loss of flow to be 3 l

limited to the design flow of the eductors, it is determined that a reduction in containment spray flow of 12s analyses. has an- l j

' insignificant effect on the transient' safety'gpm  !

Additionally,ipump performance testing-(required-by section IX, l 2WP-3230) demonstrates that the pumps deliver greater.than the j required flow of 2700 gym (assumed in' analyses) plus the 12e gym loss due to l'eakage. With the margins available in tha current analyses,.the-design limits for these parameters will not be 4

l exceeded and }do not adversely affect accident Additionally, mitigation the

capability considered in the analyses.

containment spray flows (77% degraded) used in the containment  !

analyses areiconservative with respect to the actual pump flows.

i It was concidded that no design limits would have been violated.

• z section 6.2.6.11 l NaOH Serav Additive System Denian Bases (Ref 2, I

! The NaOH spray additive system is designed to operate Section in conjunction with the containment spray system (Ref-2, 6.2.2) to remove radio-iodines from the containment atmosphere following a LOCA. The containment spray system provides the supply water ;for the spray additive eductors. The spray additive .l l

system is designed to the following; criteria: I a) Maintainl the containment spray solution pH to achieve rapid absorption of radio-iodines.with minimal caustic corrosion of materials and protective coatings within the containment, b) Maintain the containment spray system nozzle spray pH l

, betweents.5 and 11.0 until such time that a Decontamination i l

Factor {DF) of 100 is achieved. ,

c) Achieve a containment sump pH equal to or. greater than 8.5 l but less than 11.0 after all the spray chemical mixes with j the available water inventory including, RWT, Isafety i injection tanks, boric acid makeup tanks, and the reactor coolantisystem blowdown to assure retention of iodine in the sump solution.

d) Removel elemental and particulate iodines with the minimum first order removal coefficient in accordar.ca with WASH' 1329.

Todine Form First Order Ramaval Coefficient i

Elemental 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />s-2 ,

Pa ticulate 0.45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> *2 l I e) Minimize the possibility of precipitation of the spray solution within the system or its inadvertent introduction into the refueling water tank.

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FROM USNRC ST. LUCIE  !!/30/94 16:52 P.10

JPN-PSL-SENP-94-079

' REvIDICH 0 PAGE 11 OF 15 i 1

f)

' System' materials are' chosen for compatibility with sodium-

. hydroxide.

j l g) Seismic Category I, Quality Ghoup B, and function under l post-accident environmental conditions (based on location).  !

j h) Perform its function following a LOCA, assuming a single j active component failure.

.i The SER does not discuss the NaOH spray additive system since it was a backfit to Unit 1. The FSAR (Section 6.2.6.2.1, Ref 2) i provides discussion of the design considerations for the spray j 4

additive system and concludes that the system meets single '

i failure criteria.

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l 4o mustvers or enn EVENT 1 i '

The following discussions form the bases for assessment of the l

l safety consequences and implications of the as-built NaOH spray i

additive system design.

I i For the scenarios postulated below, the NaOH spray additive As

' system design does not impact the likelihood of a LoCA event.

a result of the design, only the radiological consequences As such, the

increase when considering design basis events.Lucie remains unchanged by baseline core damage frequency for St. However, leak before the NaOH spray additive design defect.

break methods which have been approved by the NRC for St. Luci,e

' (Ret 13) have shown that large break LOCA scenarios are not credible. Nevertheless, the three design bases scenarios j

described in!the FSAR are those evaluated for significance to plant operation and saiaty.

t l Damian Bamin Event - Scar.arios I t Three FSAR deeign basis events are discussed below which l

illustrate effects of the NaOH spray additive eductor piping and f how ECCS suction piping pressurization could be managed assuming l

- 128 gym leakage

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1, Larae Break LOCA With LOOP and One Failed Diesel' One train of ECCS equipment would be operating due to the failed )

diesel, one! containment spray pump would start early in the event as a result of high containment pressure. The idle ECCS suction header would be pressurized by the operating containment spray pump via the NaOH spray additive eductor piping.

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! 't .FROM USNRC ST. LUCIE  !!/30/94 16:52 P.!!

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' Inthecasebflthe.limitinglargebreakLOCA (Double Ended Cold l

Leg Guillotine) presented in' FSAR Section 15.4.1.1 CSAS would notuate ear,1,y in the injection phase and the idle ECCS suction header would:be pressurized. In this-scenario, a vent path

' exists from the idle train's suction piping through the LPSI pump to the RCS via safety injection piping following RAS. The relief valve may lift upon start of the available containment spray pump, but would be expected to resent when the available LPSI -

l pump is secured at RAS. A calculation (Ref 15) shows that the j pressure drop through the vent path to the RCS is sufficiently small such that the. safety relief valve would ressat. The j

calculation determined that for RCS/ containment backpresdure less u than 38 psi'the idle EcCS suction header would relieve through the ECCS piping to the RCS. FSAR Figure 15.4.1-27 shows that containment pressure drops below 30 psi approximately 2 minutes following the large break. In this scenario, safety relief valve i leakage would not be expected to challenge the operating ECCS-

train and the consequences of this event would not be made more ,

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! severe.

2. farae Dreak LOCA With Sumn Valve Failina to Open .

Both trains of'ECCS are initially actuated in this scenario and operate until RAS occurs. At this point one train of ECCS is i

lost as a result of a containment recirculation sump valve '

failing f.o open. The idle ECCS suction piping is then i pressuriced by the operating containment spray pump via the HaOH l spray additive eductor piping. ,

I In this scenario, RCS pressure would be sufficiently low at RAS i such that the safety relief valve would not lift. The consequences of this event would not be made more severe.

3. g3gil Break LOCA With LOOP Which Does Not Actuate Containment L Enrav  !

In the case of the limiting small break LOCA (0.1 f t ) as' 2 i

described in TSAR Section 15.3.1.3, the injection phase of the l

accident continues for approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> based upon HPSI flow rates and the RWT volume which must be injected prior to RAS.

I Control room operators use E0P-03, Loss of Coolant Accident i procedure (Ref 12) to stabilize conditions and provide core i cooling. Core cooling is accomplished by the safety injection j system or shutdown cooling in accordance with E0P-03, or by i exiting Eop-D3 so procedure op41-0410022, Shutdown Cooling.

riace decay heat would be low several hours into the event, it woead be possible to cool the core using shutdown cooling. The shutdown cooling alignment is discussed in TSAR Section 6.3. The ,

plant could enter shutdown cooling as soon as 3 1/2 hours after ~'

'i the start of the cooldown. Use of shutdown However, cooling prevents the should the .plemt I-idle train from being pressurized.

enter recirculation phase with the containment' spray pump in a piggy back dlignment to HPSI pump, the containment spray pump

• i would pressurize the idle train and result in leakage via the -

safety relief valve. At that time the ECCS sump alarm annunciates and alerts the control room operators to the leak 4 -

J FROM USNRC ST. LUCIE 11/30/94 16:53 P.12 ,

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JPN-PSL-S ENP-94-07 9 REVISICH O

(- PAGE 13 0F 151 path. Shutdown cooling system alignment would be identified as the-method for long term cooling.

Possible success paths.have been described for specific limiting rSAR design basis.LOCA events,.however the full-spectrum of LOCA events have not been evaluated. LOCA scenarios would require activation'of the on-site support centers; the Technical Support

-center (TSC),and the operational support.Conter (OSC). The following discussions present leakage identification and-strategies which could be carried out in LOCA scenarios to-mitigate the' effects of the safety-relief valve leakage:

1. 13us control room would be aware of ECCS room flooding due to relief valve leakage via ECCS-sump level alarms.and RAS radiation monitors in alarm.

2. Control room operator detection of. system leakage would occur by procedure (Step 8, EOP-03 LOCA, Ref 12). This procedure step addresses identification of a LOCA outside of containment and directs operators to locate and isolate the leak.

3. It is likely that the safety relief valve leakage would be identified early in any LOCA event by operations and/or Technical-Support Center and action taken to isolate the leak. The isolation of these safety relief valves could be accomplished by disabling the relief valves or by isolating the NaOH spray additive eductor piping.

4. Procad ral steps direct use of the LCRS to pump back leakage to containment. In LOOP scenarios, it is reasonable to assume that power would be restored to the sump pumps using jumpers or by re-energizing non-safety electrical buses in the hours that are available before flooding challenges the operating ECCS train.

5. If aafety relief valve. leakage occurred during the recirculation phase of an accident and dose rates in the RAB prohibit ECCS pump room entry, it is reasonable to assume that the root cause of the leakage could be determined and a strategy implemented to protect the operating ECCS train from flooding. The control room operators and emergency response organization would have.18 hours2.083333e-4 days <br />0.005 hours <br />2.97619e-5 weeks <br />6.849e-6 months <br /> to diagnose and remedy this situation before any active ECCS train degradation would occur.

I '

While not all possible LOCA events have been individually reviewed, the LOCA events described in the FSAR have been reviewed and it is concluced that the consequences of these.

events would have not been made more severe as a result of the NaO!! spray additive piping design.

l FROM USNRC ST. LUCIE  !!/30/94 16:54 P.13

- - _ _ _ _ _ _ _ _ _ _ _ ____ _ b

- - , -. . -. ~ . -. .-. - - -,

JPN-PSL-SENP-94-079 )

i REVISICH 0 '

! PAGE 14 OT 15

5.0 CONCLUSION

S 4 LOCA scenarios are considered low probability events for the

^

large double, ended. break or not credible, however they are '

fundamental design basis events that are accommodated in the St.

Lucie plant design. Arguments presented in this evaluation i demonstrate that the consequences of the above specific analyzed )

design basisievents would not be made more severe than FSAR' assumed analyses as a Issult of the NaOH spray additive piping design. This evaluaties also identifies the flexibility of plant design whichomuld be drawn upon in an emergency response situation toimatigate the effects of ECCS suction header ,

pressurization. ECCS would be able to accomplish its intended i function based on the events considered in this evaluation. i I 1 The results of the dose assetsment indicates that the The offsite dose to dose is less than 1/10 of 10 crR Part 100 limits.

the exclusion area boundary and the low population zona remain l

)

within a small fraction of 10 CFP Part 100 limits and the dose to j the control room would not exceed the conservative values '

described inithe FSLR.

l -

)

6.0 YERIFICATION'EDMMARY l

The scope of:this verification was to review the inputs to l deter;mine if;the results were reasonable. The method used for this verification consisted of ensuring that the applicable refs.'conces, codes, and regulatory requirements were identified and addressed. The inputs are correctly selected and applied.

• l 4

The conclusions provided are reasonable with respect to the inputs and discussions. The verifier concurs with'the Nuclear Safety Related classification of this Engineering Evaluation. The rationale in. assigning the safety classification was verified The against the requirements of JPN Quality Instructions.

verifier concurs with the conclusions outlined above.

I I

5 11/30/94 16:54 P,14 FROM.USNRC ST. LUCIE

~~~ ~ ~~ _ _ _ _ _ _ _ _ _

o

=' *

]

" JPN-PSL-SENP-94-079 REVIEZtN O p?

PAGE 15 0F 15  !

f l

7.0 REFERENCES

1. St. Lucie Unit 1 Technical Specifications, Amendment No. 129

2. St. Lucie Unit 1 FSAR, through Amendment No. 13 i

3. Initial Assessment of operability, " operability Assessment1994 of i Safety Relief Valve SR-07-1A Lifting", dated October '23, i 4

- 4. St. Lucio Action Request (STAR) 1-94100259.

O,

5. St. Lucie Unit 1 Safety Evaluation, JPN-PSL-SEMP-94-076, Rev.

" Increase of Engineered Safeguards Suction Piping Design i Pressure".

l 6. Licensee Event Report (LER)94-006, Rev. 0 i

7. Inter-Office Correspondence, JPN-SP-94-152, NaOH Cross Connection Design Review, Dated october 31, 1994.

8. St. Lucie Unit 1 Engineering Evaluation, JPN-PSL-SEFJ-94-033, Rev 6 " Assessment of Safety Consequences of a 128 GPM >

Reducta?n in. Containment Spray Flow".

9. St. Lucia Unit 1 PC/M 231-77, HaOH System, dated March 3, 1978.

10. St. Lucie Unit 1 Safety Evaluation Report (SER) and supplements 1

& 2. l

! 11. Inter-Office Correspondence, JNo-HP-94-058, Dose Assessment of Additional ESF Leakage, dated November 8, 1994.

i

' 12. Energency Operating Procedure, 1-EOP-03, Loss of Coolant Accident.

13. NRC letter to J. H. Goldberg, St. Lucie Units 1 & 2 - Application of Leak before Break Technology to Reactor Coolant System Piping i

- TAC NOS. M84560 and M84561, dated March 5, 1993.

4

14. Deleted '

15. St. Lucie Un: t 1 calculation, PSL-1FJM-94-019, Rev. O, LPSI System Pressurization Response Due To Common NaOH Spray Additive i Injection Crosstie, dated November 30, 1994. .

l

16. FPL to Raytheon Engineers & Constructors, JMM-JB-94-072, St.

Lucie Unit liNotice of Substantial Safety Hazard NaOH Spray Additive System, dated November 18, 1994.

I i

l t

l 4 FROM USNRC ST. LUCIE 11/30/94 16:55 P.15 TOTAL P.15 i

___f

. wg m FaxT jMemo ... i u g' ~~

"y .,  ! We n' p d .#p ueyr

~ l u.se. .

i - w  :

Communis

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4 i

a a

TO: ADMIN-PSLDim. Pet.-no rm No.as7-set-aans rsea tw w7ts Pea 4- - - sis.'o4 ttE saat 1

e acommar am n.ws.tur u. x.w

.-c. l f

a* s e t 'pase: = ' '

~

o.,r, mons meman  !

i

,,,, nore ,,,,, g *** '8 +

Ehde & PGfte ==

L p.o. nessee

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c. c. s, . x $s a .

e e m : 9 y, wort: Pan Bestnusmoess esa poems sta m twes .f,gg sw nos (sos

' (raci ask _ , , , , , . _

Mt.iW7sA- N ots o (soci M i M(sus dwt.M M tresi n (se/M l i nm mt u m a vsew.m.em s me w ws ro ___.

)

(.oMM akau ~Nues.we r_=-nsTE __

vamosanemoracnema sweep f-fL_ sarsw as.nassisearsos d a. __

esserrr.zas cesses A storsu museum (srsi et , oT , , , , _ , , ,

. .,,, L sarwomes insse semest ei.iaio , __ ,

i asysammes muusses tear was jdg:. (nar se. i.25th-- ' .

(mar fress 1 (ast do.1 h -9".,f4% ,

. (ser TVts) (ser so.) _ _

i persesecs commeurs (noc Trra) & (Dec so.) W+l Tid *9.21 ,L.

(D00 TTr8) _ (nac me.3 _.

I cacc Trys) (poc 30.) __ ,. .

i saumssa sucreas) --_ Ch edhuAunst omar sa.eeemeg,@*.ma.C 3m

' t h tsanigy 64metase. tA M.i; en.tt m .g L *_5_

spesassess coac n es) (see me.i .. .

serenstese at (noc Tres) .,, (Dec re.) ___. _ __

i 3

Amstt2cun3. sWTRtsan F151.5 SWTRY -.

.. . . - 'h

-,o -os. . .a. .a._

Pl 98W19m at6 ' j[ . Data RW13850 Br. SC *19272m er:

Q A RECORDS eare* ***' 8"" '

DECe5 9M pese h, Rev em L s

F ROr1 USNRC ST . ' L UC I E. 12/14/94 00113 P. t e

.,.e .

7-

' ~~

i gr-

- TO: ADMIN-PSL Tsc-13 '94 TUE 14s33 ID FPL PSL-PEG FM NO:487-G31-2200 FML1PE G716 POG

,l l

!+ l rAnfm ATION PSL-1ME 94 019 s w 0 ST. LUCIE 1881T 1

$ trf,1 SYtTDI PE550RI?4T!ct RESPONSE  :

DUt TO CC:90008 Me0H IP9ECT' ION CRossT:E  ! ar.' ,1/F'/84 l

i PL. l i ,_ i ,, la I

l e 1.0 PtlRPOSS/Scoph The purpose of this calculation is to 68te,: sine if relief valves SR-07-1A & 18 would lift with under the following conditions 1 l

' i

1) The common 140H dnje:Cien = roar. tic re~.ains in niace r_s i installed ptTlFO/M 2#1=77.

2) Large treek .f.,os:< et coolant Acc.ident .LnIncA) .

j 3) Only two of die four sets of Svfety I/doct:.on Wender valves (V3615, V3614', k V36?.71 or V3425, 136:-6, V352'i; e.e V?635, V3636, V3637 l ou V36c5, V.s446 cpd 136: 7) ha.ve ope;ied,

4) Loss of offsite Power (I0.r)

5) Los, of an F.-mroency Disoal Geherator (EDG) prior to safety injection isola-ion '.alve: ope.a.ing.

5) Recirculatic A*:tuation a sgnal hem occurrei.

i Under these condi iosa, only o:.e train af Ecc8 g ipment would h:

) operating due to the 1407 and failed ER. One cantninnent spray

pump would start early in the :.vant as a result of high

' containment pressgra due to the., LaLoch. The anciretzlation Actuation Signal a, cts to shut Gown the Low pressure safety '

In'jection (LPsI) pumps. Recire'21stion flow is provided by one High Pressure safety Injection (MPsI) pump. The idle ICCs suctic.7 l header would be pressurised by the operating containment spray pump via the NaOH ed".ctr7 piplng pa report d W 17.3 94-Cos. If there are no vent psi.hs to rwlievs. the prei.euxi.:ation of the id1-

' 2ces suction headar, the .:orrespot. ding ECCO s'.:.ctio;. relief valys ,

l (SR-07-1A or IB) Qill lift ut 50 paig t.no ::alia2 te tr.a pipa l d tunnel or Eccs pump room.

i This calculation will demonstre.te t .at under en:tse copditions th'n r flow path throughlthe idla LP81 pump to th4 5=futy InSction headers will relaeve the pressurisanion,of the idl. Eces suction header prior to lifting 84-07-:A or 15 Tha flo pm.h smalyzed it

depicted in Figuq 1.

! l 2

4 4

l J . , ,

d l -

FROM USNRC ST. LUCIE 12/14/94 08:14 P. 2

r gg

g--- * ~ '-- DEC-13

'TO: ADM I N-PSL

'*)4 TW 14:34 ID FPL PSL.-PEG FAK ND 487-631-2299 FR4.INE C716 Pee a.

I cALtm).TE0N _ eq-)i &,p.1.01s

(' ST, LUClK UN17 1 k_.h E.__

l f

t*S7 SYSTEM PRf.551AL12EDON F21*0ESE " 1130/8' Dus TO C00000N NaOH INJECTION CAOSSTIE I ou._ _. , ., m..

P L _. . i _ ....

2.21 Abasco Edector Performance Test at Ingeracil-Rand Comoany, perfors.aii pa:rch 7,1978 l

sheet SI-N-1, Revision i 2.23 s7'80-e-125,l 2.23 vn? nog e yo?oci iptnyAc. Dreving 4770-3n s, movision 5

\

2 24 General Ma$ntenar.ca Mcedure !! ,. XG210, Revision 14, Bench Testing of Safsty Relief Valves '

l '

..:s '.K.ar.t ;1n 2.25 St. Lucia Unit 1 Technical 5peci f tentf.c:.0, 3.G KJrrscDOLOGY f I and v3114, Three check valves (vo7000 or V07001, v3106'or v3107;th for the V3124, V3134 or V3144) must open to provido a flow pa Ecce sucist. debba' to -c13s.vs The first through ths :.dle objactive Lyttcalculation of tna puan to tita e s.:etf injoction of11heteemicu.]hce.ders. st th: rrrpiredSince differenui'.G nrer.3Jea very little dimeneisnal to c:cr.

the check valves l(Sectiott 5..'). 4 data regarding the check valves is available, conservative estimates will be made for some of th= di.an=1cnr.1 dat.=. Fig.aret 2 delune&*.at the FTec ECC7 Dic.Gr&M for a typicti cWing e. hock valve.

The differentia 1Ipressure recuired tai apen the ch= W *fr.17s i; W pressure requiredoftothe overcome staa hinge.the weight of the After the discvalve check and the  ;

frictional force differential preassure (Ap) is determined, the marimte pre ==.=ra '

l

' n.=T.&treme er W.s ohm 4 ve>3 vet vill be f atarminad.. N we,:est case 1

scenari: as .shan containment is at it:5 peak most-LM_Ach resirws1.tivu phal thtAse presag e cna the strong *st P, PSI puz.y 4e

• j injecting headers. to the[oontainment through only two satety in$sotionThe l

be compared i.o the set pressure of relief valves sR-07-1A's 13 (60 psig minus 34 tolo w ee) tv dst.nr. i.n '.,': W whwak valves the reliat vaaves lift.

will op.en befor I

If theipressure required to open the caeck valves from section'r*.e' i

! 5.1 is' floss than! 60 psig, the next step will b, te detemi: .c peak prassure in' the EccJ w.wnian h.=dar when the NaOH cross-tie is delivering its maximum flow to the suction heaMr (set tden 5.2).

The Naott flow will be required tG Lwlieve through the associa(ted LPSI Dump and the safety injection headers to the RCs.

The pressure less between the accs section and Lt.e Ec.s with the worst case fluid flow, the elevation head bettXc4 thc. RCs and the ECCS suction, and the containment at its peak pett-TECH recirculation phase pressure will be determined. The ranults will

• be combined andIcompared to the set pressure of relief va3ves SR-07-lh & 15 (80 psig minus at tolerance) to determine if tt!n t

relief,velves will open.

t f

I t 1 l h FROM USNRC S . LUCIE 12/14/94 08:15 P. 3

.- - 2 -_ .

Y To: ADMIN-PSL DEC-13 '94 T1K 14835 ID: FPL' PKi..@E3

• I FAX bC:dB7-491-22M FA4.IbE/.716 P99

.j J

, A CALCULATION __ PSL-INN-94 019 __

4

$7. LUCIE UNIT 1 ' 0 __ =

LPs! SYSTEN PRESSURIZATION RESPONSE

" 11130/84 rm to cmmmet NaOH INJECTION cactstIE l b  ! r_ 4- ., la 1, l

! I 4.0 AssuuPTrops/aASE;s i

• I j l 41 ~

The weight effective; acting;e.t weight t.heci'center the disc of /he.~ly.

cravi essembly (cog) of the is une 41sc.

This is consistent with the methodology used to calculate i the: velocity required to fully open a aving check valve per Ref, 2.3.

4

! 4.3 As the cantar of eravity of th-e disc arm assembly isn't known, tas aftsetivt weight of a swing casak val-e is' assumed so act halfway tetwee. the center or Las valve cire and. its butt.o% edge. Tite ef:!ective weight ist a soerrination

.:. the disc &pe tus hings arm. The v&ight c tha hinge ars '

woald 3me I;he COG of the umut.y shova the etnter of the l disc, therafsts this asmuzution is concc: .ativ=.

)

I l 4.3 An ispir.;.. rant =gle is not availabic fer V07000 L V07001.

An 1::pingement angle (8) 02 25' will be used !=r 707000 ;

ts _ . . . .. 4 .. w..... a _. e* 4s tha sans I

..g;pgg, vahv as una =[g h ar V3114 - 73244 and V3106 & V3107. The smaller valves backseat against the valve body (msf.'s 2.S &

1 3.6) which armat:er. a larger A angle than valves such as 70700G & V07001 which backsoat against a stop extending from the valve. =cvsr (Ref. 2.23) . Since # = a + $ and # is san 11er for vo7000 A Vo7001, it in conservative to use a f of as'.-

t 4.4 The, ===ffi=1:nt =f friction (;:) itmesur.edtobec02h.d..1=h 1s based upon the highese iW trom Table 15-4, Ref. 264.

l  !

! 4 . r. suoyancy forces on the disc are negligible.

4.6 Thedistancehatweenthecenterofaswingcheckvalvedisc to the hinge f.1r. fcc the valves ev.alustas. is assumed to be i

eg.=1 te the d * ---*- ~* ***h j i 47 The hinge p n{dir.acter (d) for V3114, V3134, V0134 V:1144 Et j u is assumed is be 1.375". Ths.m asranption is concervative hased upon tb3 known hirgo pi.n diameter ter V32.os c v31c7

, {Ref 2.3) which le 1.375". V3106 & V3107 are 10" check vs.\ves with R.' disc weight or 33.34 lbs., wheresa V3114, l 1

V1191 V313a!& V31dd are 6" chmak valves wit:h a di ne semi m+ '

! of 14'.3 lbs. .refore, 1.375= is conearvative.

l Am Impingement ebgles of 25' sto given for V3114, v33.24, V3134 i 6 V3148 and V.3105 4 V3107 (WAf. ' A 2 1. 4 E di and asn'.ut**i for l V07000 & vn1GG1 (Assumpt;i.uu 4.3) . Th4 impingement;. angle (#)

of 'owing chedk valves (Ref. 2.3) la defismed se a r # (See i Tideaw 31. Itd is es6ueed that a = 8/2. =laiuh i= con ==rvati w=

l J for the check volves eva?.unted in this calculation because _

I t

j FROM USNRC ST. LUCIE 12/14/94 08:15 P. 4 l l i l

'~

.i F.

TO: ADMIN-PSL LEC-13 's4 TLE 14a J6 L !Dt FPL ." MIL-PEB -

FM NOs 487-691-2200 FMLINE 88716 P10 I i CI -

I , l_ calcist.ATION PSL-1FJN-94.019

! ST. LUCIE UNIT 1 a. O I i LPs! $YSTDi M!8.5MMItaving afspossE lDWE TO CO 6 Ha0N INJECTION Ct.0$$Tgg on, 11/30/94 b I em .s s la l

1 fros a r=*tiew =? the valve drnwings, # ist much largeir.tha.n a (Set,'s 2.5,i2.6 & 2.23).,

I 4.s esembly offecti'.e weight ue=d for vofooo 4 vo?001 The wiu hi. misc j/ ,9 X 33. 34 lbs . - 46.7 1he. *f. 3 4 1ha . is the vrfret of 10" valves V3106 & V3107 and /u . is tha :rstic of diameters ofi tne 14" V07000 & V07001 to the 10"e**volve.s. valves This ic ==r.strvative basse unen a cooperison or the V1M4 - V3144 to the 10" v3 Aus s Viis7. e V3114 = V3144 valve discs keigh 28.32 lbs. A ratio of ~~p/s. X 38 32 lbs. =

47.2 lbs. Then actuni weight of V3106 4 V31ci is 33.34.lbs.

I 4 10 TM hinge pin diameter (d md for VO7000 & vo7001 is t;gsw.ad to bei,3". This assumptien is reasonable based upon tha 'Lacsn hin"e : pin diameter fr:r V3106 6 VM.07 (def. 3 3) which is 1.375". The hinge dinmater .ia sai:7ated bas.e2 upon tt.4 uma of t=c pin tesisting ct.orr 2 r.ds, marercre, the ralationshipl)gaeveen (i.e., i d ,. , * (onus) thp. Using pirt (15suotesw the ratic cf is the a square assumed functLon weight. "ci' sl14" chsch valvet diac (5d.) 1ha.) to the weight of a 14" nbar',,st valve disc ( 34 . 34 lbs . ) , then !

.2..

(%,ress) = "* P,. ,., X (1. 375")' = 2. 57 in.8 I I l '

l 6 ,,,, = ( 2 57 in. ') = 1. 6 0 in .

i 1

! i Therefore, 2a is conservative.

l l

, 4.11 'rt.e affective disc diamatar (D) = 1.1 X D. pn Phere D,m m,.

!. = valve nominal size. This is consistent with the "

mecnodology Est 2.2.

l l d .12 The maximum lconta'inment pressura during recirculation is 17.0 psig var Raf. 3.13, Fi;*. = 5.2-lh. This is e.lse 'ths j ,

nressure m'.,Itna RCS injection nossles, t

i j

l 4.13 Par 7.af. 2.20 the !3soH edr.auru have 11/16" crifices. Per

~

j hf. 2.21, the maximum discharge flowrate of a 11/14" I s eririca =dustrne was 1G3 spm. Fsr conesvatias. ths Eductor i disas.rgt flouree is Gscram' to Dr. No 7p with 100 gym rioving throug'u mach s&fsty injs= tion hcadar.

I 4.14 The us.d.Tum l lone ps! pury) fi m te occurs when the i stronevast ugsI pung le runnin- all tur nr.isty !.n'. action.

' linse a: a ebn and. there is zero RCS pressure. Per Ref. 2.8.

this finwrate is 133 siri.=. This f1=.tre.:: :, d=utisd (066.gir.E)

! ximate the. flowrate'through just one line whdn two to!

ol' pfour!safatty t e injection lines are isolated. 366 gym will be used as the fl:r.: from th= RPSI eystem in ,

i calculations of ps cTee ?.one for the piping downstream of 4

t 8 i

FROM USNRC ST. LUCIE 12/14/94 08:16 P. 5

4-- TO:ADM

-DCC-13 ' 94 Tts I N-PKarPL 14 :35 ID PSL-PEG .

FM NoldF1-691-2200 FMLIFEjit716 P11 +

l 7 8'AMUL4'!!0N PSL-1FJN-94-019 t

.l

' O Es" *E l

I St. LUCIE latIT 1 i LPi! SYSTDI PRrJ.5'JRIZATIM RESPONSE DUF T0 Colge001 Ma0H INJECTION chussTIE 8 "-

l-w b 11/20/:;4 5,

w i

i l tha'intersanction sf' the LPHI and EPSl' t'.:tlas. Thist i:1 conssrvative because a) the flowrate frce the IDFsI pump wouM dovreans with incraastac 20s prassure and/or flow from ttW LF51 hosther; and 's) t$ade relatic;.r.hi-to blowrmte .ts an exponential function (1.cf flew rees. , flevratptamne equared is ptoportional to flow resistance). Therefore, using 3s4 gpn maximises the pressure losses.

t I

a.15riiejacasityofvsterusedwillbeatatemperatureor550r.

Tr,e. lowest rafueling water tank teerparature is 55'F per Technical Spoolfleation 3.3.4 for Modes 1 & 3 (Maf. 2 25).

h; densitf Of water increases with decreasing two.natraturm.

Thatefore, the weight of a water so'ur.n I is higher et p. 3 war tmeyers.ture pad using densit.y at 50 F is consar-rativa. Tr.e I

conversion of feet of water column to poundb per squsra inch 8 (pui) its Ger{wity(p) E So*F = 6.S . 4 lbs. /ft.~~

N'refore, St. H,0 *A (42.4 lbs. /ft.') / (14 4 in.*/ tt.')

l I

0 43 psi /it. ,

l l

4.16 Me ha,ad loss is assumed through the Li52 puany becauss thess 3 pumps use hiO capacity pumps (design flowraw - 2000 gpo i par Inst. 2.12 - Table 6.3-2) and the annumen :fAowrate; l

G weugh the pumps is low (200 gym).

j l 1 l.

2e 'aEUI.hTION a.

i 5.1 DITIJ'EREFTXN., raRSSURE E3952 RED 20 0?Ett LPAE i M Vffit YP1VES TO HIDTZDE 3GGe SUCTI0li 7ENT PATE I

3.1.1

1. CAtcULA,TE RitQUIRED AP To DiaBW CHE 3 VRIfiB3 [

t V31'4 4, V3124. V3134 & V3144 k

I Fidur: 'e dep4e'ta the Free Body Dia$ ram for a sw3.nq crignox veivs stuch 52 V3114, V31?ft V3134 Ei V3144. From Ret. 2.1, tM fonswing parametere are known 9 l

i i i, ~u~ mlpniyame;.t Angle - 25*

w - mitmot wo wetsut = :8.2 lb:-

h i 4

4 pi.so : Diameter = 6.8"

, z.m _1n, ,=.=-. -

a m. 25'/2 = 12.5* (Assumip. 4.8)

L ..._ ._. _

1 -

.G FR0t1 USHRC ST. LUCIE 12/14/94 08:17 P. 6

TO: ADMIN-PSLIDeFPL PSL-FED

• ~; 10*%;-13 'N T12 idade FfW NO 407-691-22GG FAXLINE 4t716 P17 i

i l CALCULATION Pst-1Fh-94-019 . , _

ST. LUCIE UN!Y 1 w 0 LPst SYSTEN PRES $URIZATION RESPON5E

,.p 11/so/s4 _,

% in4 70 7,W113 Hec:' '!'KOT!PP C*0?mE I l

== i l. ,,,,',_,,, __

b =. 1.1 s 'i . _

Tho' allowsere .et pressure t.cierance :!er su-07-14 a is 1, 3g par Raf. 2.24. Adjusting the set pressure for toleranoa:

i (60 x 0.03) = 58.3 psig P,g = reo ps.4 -

Therefore, the margin between roller vatves &E-G7-3.A & .:; .

set, pressuc*. a'ra l'he pa scret-e e e;ve e.e?x vatver iss M&AGIN = Fatt * *E 1. - P,ms HARGIN = 55 2 psig + 4.07 pai - 36.59 psi = 25.65 psig l

• ".52cf4th

,. , sheek valves V3814, V3134, V31314 & V3144, y4A06 &

Yss,o? and Tovese a ' ire';nor. etl.11 et;eas :sciers k-uT-it .', it will lift.

s.2 ramt 73358055 3N 9508 3005 EEhDER W2TE unas yg,ggr mEd t

wmm r.re raur so ess mas i i i S.2.L des.WT ATIO!! CF PRF.SSURE I458 IN SAFETJ UlJMM61r StauG ;[/ ROM INTEZ22t"!r.'M is? UFI/12sr % IE'E UG f WITM 1C0 GFM FR006 LPSI AND 266 FROM GrM RPSI t .

l Trs- f r.e'erition 4 14, the maximum HPSI flowrate through the I pipity asbtut.ag thr. f.!:';arsection of the LPSE/NPSI systems le 3h. 3;* (-/a ' sere f .o** from the LPSI syst ) .1 Par Assumption 4,13, the tswrate AL'lm *).n MeC*l etw: ors: thectomt &

relieved through each safety injection hamaer is 1u0.gpa. A The combined flow is:

' a-Oseng, = 365 Upa + 100 gpa = 466 gym U$::c 4 +.::6 C Lv raisistaact fe1*org (M) framRef.2.17fthe pressure loss in the WA st can:as pipe;s (piptas 14 and. .L5 from Attachment 1, Page 1 - Raf. 2.17) can be calculated (The maximum "K"-Factor is the 6" line that will produce the '

f etsimum A?)t g (ma*. 5.11) l Pina 14. f6"isen. inGi2 Ku 10.0" Area (Ap) sm 0.1469 ft.' (Ref. 2.12) l

= 1. 04 ft.*/ 7m3 4 q f 4G0/ (7."l.!) ( An) '

y '= ' g", 0 4 *t. '/.g eq'; /10.14 69 f t.8) = 7 07 ft./sec I

l n = EE/29 hd = fU...'5)(/.9i fi.. / s.4 */:: ( 2 2. 2 *t . / ")

t  !

l i FROM USNRC ST. LUCIE 12/14/94 08:17 p, 7 L - .

' ~

I TO: ADMIN-PSL DEC-13 '04 71.E 14:42 IDsFPL PSL-PEG r FM NDsdW7-691-2288 FAMLI E 88716 P21 4

' \

' * -l GALN#3710N PSL-1FJN-94-019 1 4

- 0 4 ST. UM1f WIT 1 i LPl! SY3 TEN PRtJ5$ MIZATION RESPONSE - -- 11/20/94 BUE TO ColeIINI Nr,3f INJECT!all Ca05871E

- , , - =

16 .n la i

t hoogg = 1.69 psi + 11.61 psi + 5,6 psi + 17 Paig = 35.9 petig i 5.2.4 DETERMINATION OF MAF.CIM BZ*,EWEEN RELTEF VALVES gg-r:7-1A

& in ERf PRESEURE AN"4 THE 14 ESSURE TO OPDI cNYC1r VALVES Thh allowable set pressure tolerance ist SR-O're & sa ja 34 per Ref. 2.24. Adjusting thu sat pressurs for scimrance t P.. = 6 0 p s ilg - (60 x 0.03) = 58.2 psig Therefore, b margin between relief velves sp-07-1A & 13 set pressure and the maximar. ECC8 suction pressure is:

MnaGIN = 88.2 peig - 35,9 pe.ig = 33 3 psig '

l Therefore, sR-07=3a & ES will not list. .

s.o masWLTS .

This ca'ioulation demonstrates tbn with maxiuam conNinsa'st pressura during reoirculation, one RysI pump running, noth LpsI pumps i!!1a, one pontainmast spray pugp running, two safety in$ action headers isolated, two safety injection im dmes ln-  ;

service'and one Adle ECC5 train t

1) Thu upstre&R pressure required to open check valvse vans, V3'124, V3134 & V3144, V3106 & V.11C'T and V07000 & V07003. As 36,59 psig. Adjusting this value suo to elevation diffau;ance with respecy, to SR-07-1A & IB

• 32.06. This pressure' is less "

than the set, pre.maure of relier valves SR-07-1A & it (sinus 4 3% tolerance) = 88.2 psig. Therefore, these check valves will open to provide a vent p$th for the ECCs suction piping to containment.

2) The pressure developed in the ECC8 suction piping at SR-07=

1A & 13 with a vont path established througn a LPsI pump to thie RCS is 35.9 psig. This pressura is less than the set pfessure ofjrelief valves SR-07-1A E 1B (einus 34 tolerance)

= '58.2 psig. Therefore, this vant path for the ECC5 ' suction piping will' prevent SR-07-1A. & 18 from liftittg.

g I

3 I

e  :

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~M-11 FROM USNRC ST. LUCIE 12/14/94 08:19 P. 9 TOTAL P. 9

~ - NOV-23 '94 tED 12:17 IDsFPL PSL-PEG FAX NOs407-b91-22dB FAXLitE #578 P02 "

fu ,

P.o.sn asso. A= sewn.n suas eas 19 FPL. JMM-J8-94-072 November 18, 1994 RAYTHEON ENGINEERS & CONSTRUCTORS 759 South Federal Highway Stuart, FL 34994 ATTENTION: Mr. C. W.' Bailey, Pro. ject Engineer l

ST. LUCIE PLANT UNIT 1 Notice of Substantial Safety Hazard NaOH Spray Additive System j PC/M 231-77 l BFI 125-1. CWO-A-125-1 1

! FILE: SPEG 94-043-11 l

Gentlemen:

Florida Power & Light has recently discovered a design deficiency in the NaOH Spray t'dditive System. This system was installed under CWO-A-125,1 (PC/M 231-77) during a St. Lucie Unit I refueling outage in 1978. This system was designed and installed as a nuclear safety related modification with the requirements of Title 10 of.the l

! Code of Federal Regulations. Part 21 applicable.

l On October 20, 1994, St. Lucie Unit I was in Mode 1 operating at 100% power.

Differential pressure stesting of a motor operated valve resulted in the lifting of a safety relief valve on the suction of the emergency core cooling system (ECCS) )

supply header. An initial assessment of operability determined that this safety t relief valve could lift under certain accident conditions and result in containment {s sump inventory loss to the reactor auxiliary building in excess of ECCS external system design leakage.

l The NaOH spray additive eductor piping was routed through a common header which results'in cross train pressurization of an idle ECCS train when the other train is operating. The pressurization of the idle ECCS suction header caused the safety relief valve to lift. Under these conditions, relief valve leakage greatly exceeds

)

l ECCS external system leakage limits listed in the Final Safety Analysis Report (FSAR). This condition results in significant reactor auxiliary building flooding which has the potential to adversely impact accident mitigation capability. Florida

! Power & Light has reported to the NRC via the Licensee Event Report (LER) process i l that a substantial safety hazard may have existed. 1 Flow diagrams presented to Florida Power & Light depicted NaOH eductor piping independent with no common header. During detailed design, the eductor piping was routed through a common header for design and construction convenience. This change i in the flow path design was not reviewed with respect to ECCS design bases l l requirements for train separation. l

-}

P ,

1 d.0 Mt.U.eepweepe#V

l

~~~*

!OJ-23 '94 LED 12:18 ID FPL PSL-PEG FAXto 407-691-2200 FA>2 ItE **S78 P03 Page 2 of 2 A more detailed description of the defect is provided in St. Lucie Unit 1 LER 94-006.

This design deficiency may have 10 CFR Part.21 significance and should be reviewed '

by your company for corrective actions.

Very truly yours, b N0f B. D. Guilbeault Manager Nuclear Materials Management M

CJB/

c- 4 J. H. Goldberg EX/JB D. A. Sager - MGMT/PSL W. H. Bohlke - JPN/J8 D. J. Denver - JPN/PSL S. A. Valdes TEC/PSL J. E. Geiger - JNA/JB R. E. Dawson - LIC/PSL D. M. Wolf - JPN/JB J. A. Porter - JPN/JB C. L. Schaeffer - JCM/JB J

t e

i

)

~~

~~ -.

~~~~

!OA-23 '94 LJED 12:18 IDIFFL PSL-PEG FAX NO 407-691-2200 FA>2.ltE 8578 P03 l

Page 2 of 2 A more detailed description of the defect is provided in St. Lucie Unit 1 LER 94-006.

This design deficiency may have 10 CFR Part.21 significance and should be reviewed by your company for corrective actions.

Very truly yours,

.b N0f B. D. Guilbeault Manager Nuclear Materials Management Y

CJB/

c+ w J. H. Goldberg - EX/JB D. A. Sager - MGMT/PSL '

W. H. Bohlke - JFH/JB D. J. Denver - JPN/PSL S. A. Valdes - TEC/PSL i

J. E. Geiger - JNA/JB i R. E. Dawson - LIC/PSL D. M. Wolf - JPN/JB J. A. Porter - JPN/JB ~

C. L. Schaeffer - JCWJB 4

S

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