ML20237L689

Proposed Tech Specs,Revising Table 3.3.3-3 Re HPCI Response Time

ML20237L689
Person / Time
Site:	Hope Creek
Issue date:	08/13/1987
From:	Public Service Enterprise Group
To:
Shared Package
ML20237L476	List: ML20237L474 ML20237L689
References
LCR-87-02, LCR-87-2, NUDOCS 8708200301
Download: ML20237L689 (21)

Text

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ATTACHNENT 1 l

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l PROPOSR3 CHANGE TO TECHNICAL SPECIFICATIONS HOPE CkEEK GENERATING STATION' I e . FACILITY OPERATING LICENSE NPF-57 DOCKET NO. 50-354 LCR 87-02 i

I. Description of the' Chance l

' Revise Technical Specification Table 3.3.3-3, Emergency Core Cooling System Response Times, for Item No. 4 High Pressure Coolant 'Inj ection System ( HPCI), from less than or equal to 27 seconds to -less. than or equal to 35 seconds ( see l

~F Attachment 2 for a marked-up copy of the af fected page) .  !

This change will provide Hope Creek Generating Station l

( HCGS) with . flexibility in meeting the requirement for a i HPCI cold quick start time without excessive system tuning.

II. Reason for the Chance During.the Power Ascension Testing Program ( P ATP) at HCGS, test engineers 1 encountered difficulty in meeting the response time in Technical Specification Table 3.3.3-3 for the High Pressure Coolant Injection ( HPCI) system - 27 seconds. .Although the HPCI flow controller provided good l flow and speed control during a transient cold quick start, the time to rated. flow was consistently longer than 27 i seconds. In order to pass the test, and Technical Specification criteria, test engineers were required to

' finely tune the HPCI system and through extensive efforts ,

eventually brought the response time to within 27 seconds.  !

As a result, Public Service Electric and Gas Company ( PSE&G) has investigated the bases of the HPCI' system response time 3 and as shown below concludes that a change in the response '

time to 35 seconds from 27 seconds is j ustifiable.

Discussions with General Electric ( GE) have revealed that i rather conservative design assucnptions were utilised in

~ determining the HPCI system response time. These assumptions include multiple equipment failures and the lack of availability of the Automatic Depressurization System

( ADS) . GE has indicated that if more realistic criteria are employed, such as an automatic HPCI initiation at Level 2

' ( L2) during an Anticipated Transient Without Scram ( ATHS)  !

event and maintaining the reactor water level above the top of active fuel, an increased HPCI system response time is justified. -Therefore, PSE&G proposes, with the concurrence df General Electric ( see Attachment 4), modifying the design .

requirements'for HPCI system response time to less than or I equal to 35 seconds.

8708200301 870013 Page 1 l PDR ADOCK 05000354 P PDR

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I III. Significant Hazards Consideration l

This proposed change to the HCGS Tnchnical Specifications l

a) Does not involve a significant increase in the probability or consequences of an accident pr e vi ously evaluated. The limiting single failure analysis for all l break sizes and locations is the A Channel DC source failure which eliminates the HPCI system plus one diesel generator, le a vi ng the three Low Pressure Coolant Inj ection ( LPCI) loops, one Core Spray ( CS) loop and the Automatic ,

Depressurization System ( ADS) available. Therefore, this I change does not affect the most limiting design basis '

accident ( DB A) for HCGS. l l

For those series of accident scenerlos where the HPCI system i is operating, there are at least as many low pressure j Emergency Core Cooling System ( ECCS) available (see FSAR l Table 6. 3 -6) as identified above, thus a startup delay in )

the HPCI system will not become a limiting event. GE has indicated that the delay in HPCI system initiation may result in a small Peak Cladding Temperature ( PCT) increase for small break Loss of Coolant Accidents ( LOC As) where HPCI is operating; however, the increase is again bounded by the PCT associated with the limiting DB A. Attachment 3 contains revisions to HCGS FSAR Section 6. 3; however, these revisions simply support the change identified and hence Chapter 6.3 analyses are not affected by this change.

For the loss of feedwater transient, both the HPCI system and the Reactor Core Isolation Cooling ( RCIC) system provide high pressure makeup to the vessel. Both the HPCI and RCIC systems are designed to individually maintain vessel water level above Level 1 ( L1) during the e ve nt,. Since the HPCI system flow capacity ( 5600 gpm) is larger than the RCIC system flow capacity ( 600 gpm) by nearly an crder of magnitude, increasing the time to rated flow for the HPCI system will not affect the ability to turn the water level around before Li is reached. In addition, GE has indicated that the Minimum Critical Power Ratio ( HCPR) occurs prior to the initiation of HPCI and hence, the MCPR will not be affected by a change in the startup of the HPCI system.

Attachment 3 also contains revisions to FSAR Sections 15. 2 and 15.6; however, these revisions simply support the change identified and hence the Chapter 15 transient analyses are unaffected by this change.

The impact on the ATWS transient is insignificant because, although the change in HPCI system initiation may alter the time water reaches the vessel, water level will still be maintained above the top of active fuel.

Page 2

b) Does not create the possibility of a new or different kind of accident than any accident previously evaluated. The: proposed change only results in a small delay in the initiation of the HPCI system. The HPCI system will still adequately perform its required functions and thus not create the possibility for a new or unanalysed accident or transient.

c) Does not involve a significant reduction in a margin of safety. The basis for the Technical Specifications on HPCI system startup ensures that ( i) the calculated PCTs for small break LOCAs remain below the limits identified in 10CPR50.46 and (ii) the reactor vessel water level can be maintained above L1 for a loss of feedwater transient. The proposed delay in HPCI system initiation has no impact on the calculated PCT for the li mi ti ng small break events and no impact on the MCPR for a loss of feedwater transient as discussed above. In addition, the HPCI system is still able to maintain' vessel water level above desired elevations for both the loss of feedwater and ATWS transients. Thus, this change does not significantly reduce the margin of safety in any Technical Specification.

Based upon the above evaluation, PSE&G has determined that this changu does not involve a significant hasseds consideration.

I V. References Attachment 4 - General Electric Field Deviation Disposition Request ( FDDR) KT1-1691, Revision i dated October 28, 1986 (8 pages).

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ATTACHNENT 2 l

TABLE 3.3.3-3 .

EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES  % -

ECCS _ RESPONSE TIME (Seconds) l

1. CORE SPRAY SYSTEM $ 27

2. LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM $ 40

3. _ AUTOMATIC DEPRESSURIZATION SYSTEM NA

4. HIGH PRESSURE COOLANT INJECTION SYSTEN i 27 < [,g gg

5. LOSS OF POWER NA )

(

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( HOPE CREEK 3/4 3-38

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ATTACHMENT 3 i,

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HCGS FSAR 01/86 in'the suppression chamber with the HPCI turbine exhaust line.

This mitigates the effects of water from the suppression pool being drawn into the HPCI turbine exhaust line. The isolation valves in this vacuum breaker line isolate automatically, via a combination of low reactor vessel pressure and high drywell pressure.

Startup of the HPCI system is completely independent of ac power.

Only de power from the station battery and steam extracted from the nuclear boiler system are necessary.

iOb The HPCI controls automatically start the system and bring it to design flow rate withinNIEfEeconds from receipt of a reactor pressure vessel (RPV) low water level signal or a primary containment (drywell) high pressure signal. Refer to Chapter 15 for more analysis details.

The HPCI turbine is shut down automatically by any one of the following signals:

a. HPCI turbine overspeed - preventing damage to the turbine

b. RPV high water level - indicating that core cooling requirements are satisfied

c. HPCI pump low suction pressure - preventing damage to the pump due to loss of flow

d. HPCI turbine exhaust high pressure - indicating a i turbine or turbine control malfunction.

If an initiation signal is received after the turbine is shut down, the system restarts automatically if no shutdown signal exists.

'Because the steam supply line to the HPCI turbine is part of the-RCPB, certain signals automatically isolate this line and thereby shut down the HPCI turbine. However, automatic depressurization and the low pressure injection systems of the ECCS act as backup.

Automatic shutoff of the steam supply to the HPCI turbine does not negate the ability of the ECCS to satisfy its safety 6.3-13 Amendment 14

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,- r HCGS FSAR 7/85 'l q TABLE 6.3-1 (Page.1 of 2)

OPERATIONAL SEQUENCE OF' EMERGENCY CORE COOLING SYSTEM FOR DESIGN BASIS ACCIDENT (2) l i

Time (s) Events

~

Approx. 0 . Design' basis 1oss-of-coolant accident is. assumed to l l start; offsite power is assumed to be' lost, t Approx. O Drywell high drywell'pressureca) and reactor vessel .

+ low water level (level 3) are reachedi All SDGs- )

are signaled to start, reactor scram'is initiated, and.HPCI, Core Spray, and LPCI receive ~the-first.

signal'to start on drywell;high pressure.

Approxi 2 -Reactor. low-low water'1evel'(level.2) is reached. l HPCI receives the second signal to start. The HPCI y injection valve is signaled to open.

' Approx. S' Reactor low-low-low water? level (level 1) is reached.. The second~ signal to start LPCI and. Core i

. Spray is given. The auto-depressurization sequence l begins MSIVs are signaled to close.

Approx. 15 All SDGs'are ready to load. Energizing of the Core-Spray and RHR (LPCI). pump motors begins.

Approx.122 Reactor low pressure is reached. The Core Spray  !

injection valves receive the pressure permissive j

. signal to open. .

Approx. 24 LPCI. injection valves receive pressure permissive  !

i signal.to open.

The HPCI injection valve is open and the pump is at l Approx /093 design flow, which completes the HPCI startup.

bgg--

1 Approx. 34 The Core Spray pumps'are at rated flow and the Core Spray' injection valves are open, which completes the Core Spray system startup.

Approx. 48 The LPCI pumps are at rated flow and the injection valves are open, which completes the LPCI system

- startup.

See Figure The Core :is ef fectively reflooded, assuming the l 6.3 worst single failure; heatup is terminated.

>10 min The operator shifts to containment cooling.

l Amendment 11 l 4

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HCGS FSAR 01/86  ;

TABLE 6.3-2 (cont) Page 3 of 4 variable Value  !

, Maxtmum allowed delay time . 27 seconds (core spray system) j L 35' tr:m tn:ttattng signal to '-B seconds (HPCI system) j 1 rated flow available and i i injection valve wide open HPCI flow rate 3000 gpm (maximum) injected through the core 2000 gpm (minimum) spray sparger-Automatic Depressurization System l Total' number of relief 5 valves installed with ADS function Number of ADS valves 4 i

used in analysis j Total minimum flow capacity 3.2 x 10* lbm/h, at for 4 valves, at a vessel 1125 psig i pressure ADS timer initiating signals a) Low water level, and 1.0 feet above top of active fuel high drywell pressure, 2.0 psig(2) and a signal that at least one 145 psig (not modeled)

RHR pump or one core spray system is running at a pump discharge pressure or

~

b) Low water level, and 1.0 feet above top of active fuel high-drywell-pressure 6 minutes from initiating bypass timer timed out signal and a signal that at least 145 psig (not modeled) one RHR pump or one core spray system is running at a pump discharge Amendment 14 l

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, HCGS FSAR 1/85 l  ;

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, l TABLE 15.2-10 l SEQUENCE OF EVENTS FOR LOSS OF ALL FEEDWATER FLOW  ;

(FIGURE 15.2-8) l l

+,

Time, s Event 0 Trip of all feedwater pumps initiated.  !

5.0 Feedwater flow decays to zero.

8.2 Vessel water level, L3, trip initiates scram trip.

21.6 Vessel water level, L2, trip initiates recirculation pump system trip.

21.6 Vessel, water level, L2, trip initiates HPCI and RCIC

-system operations (not simulated).

51.6  ::-: = :.70 RCIC flo enter 3the vessel

=

{546 HPCI Fla0 GOTEBS TM6 VECSGL )

Amendment 9 l

HCGS FSAR 5/85 l TABLE 15.6-7 SEQUENCE OF EVENTS FOR A STEAM LINE BREAK OUTSIDE PRIMARY CONTAINMENT l _,

Approximate Time, s Event 0 Break of one main steam line outside primary containment Approx. 0.5 High steam line flow signal initiates closure of l

MSIVs

<1 Reactor begins to scram l

55.5 MSIVs fully closed l

[ Approx. 27 s I RCIC and HPCI wouldln:/c initintet on low water ggeggyg Ag level, L2 (RCIC considered unavailable and HPCI gngng y assumed disabled by channel A de power source failure).

Approx. 30 SRVs open upon high vessel pressure. The valves then open and close to maintain vessel pressure at approximately 1000 psi.

Approx. 90 Reactor water level above core begins to drop slowly due to the loss of steam through the SRVs.

Reactor pressure remains at approximately 1000 psi.

Approx. 490 ADS receives a signal to initiate on low water level, L1; ADS high-drywell-pressure bypass timer started.

Approx. 970 All ADS timer's time delays are completed; ADS valves are actuated; rapid depressurization of vessel initiated.

/ rox. 1215 Low-pressure ECCS systems begin injection with reactor fuel partially uncovered.

Approx. 1290 Core reflooded and clad temperature heatup terminated; no fuel rod failure.

r Amendment 10 l I

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HCGS FSAR

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TABLE 15.6-22 l

SEQUENCE OF EVENTS FOR FEEDWATFR LINE BREAK OUTSIDE-CONTAINMENT Time, __s Event. NKS 40 0 One feedwater line breaks wmA700 SGSL;

>0 Feedwater line check valves isolate the reacter f rom ,/ l z the break /

lhe 530 At low-low reactor water level, - RCIC \ HPCI.f P and r c a c t o:^. l iccr r initict:1 and the recirculation pumps trip. At 71ow-low-low reactor water level, MSIV closure initiates 120 The SRVs open and close to maintain the l reactor vessel pressure at approximately 1100 psig

>3600 Normal reactor cooldown procedure established.

_a5 RE4CroR sceam co tou3 uo4 Tee tsust) i

ATTACHMENT 4 I

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o 408 925 5946 Oct 29 86 18:36 P.02 GENERAL' ELECTRIC TEL.No.

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ofVIAi6ch oESCRIPYloN This FDDR is a supplement to FDDR KT1591 Rev. 4.

The HPCJ system la provided with a Ba!!*y -721 Flow ' Controller at Hope Creek. Other sites have used a Bailey -701 Flow Controller. The HPCI controller (E41-R600) the veseet at Hope Creek leaves saturation imradietely upon a &w increase durin

' ' injection event which causes the controller to take over flow control Pier to the control Aw reaching'setpoint. This provides an excellent means of initial and speed control on a transient oold quick start but delays the thne to rated flow I T M 8t'ece m **t'ct 8 N

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'g g succesTao oisPosmoN Because the present flow controller (h41-R600) provides such excellent flow and spect 1[' control during the initial transient on a cold quick start;

1) San Jose is to provide analysis to change time to rated bw to less than or equal to 35 seconds. Prov!de justification of new time with this FDDR. .

m (Continued on Page 2) ,,,

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FINAL DetroliTioN O See 4he ottuched letter From DA Hamon to 3,k.sa k

' tidedibn GRO Mod,fy caph 3.2.5 of the Sye Gek Hi h Prosure CoJcmt os Shown 2.) SystenPorofy Cks Specifi co%n Cbta Shed CDb') 2t/%B37,98 on shu.t 7 Svste in the Stadup Tot M the level I criteria ihr the H Hect$, sten sw r-up t.W .

S.') spec 4 ation atA9137 to re.ht a 35 s oi a.0 _

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GENERAL ELECTRIC

,' PDDM KT1-1691, Revision 1 Page 1 of 1 DEVIAtt0N DESCRIPTION (Continued)

Present criteria le for time to rated flow in less than or equal to 17.0 seconds.

Other sites with the Ba!!ey foi flow controller achieved rated flow in 21-22 seconds.

t g (5*1 5000BSTED DISPOSITION (Continued)- p, . : -

4n, Q 8&n Jose shall also provide confir. nut!on that increased time la jus $, fled

% kbecau 5,g

2) lo- n. g of Ba!!ay -721 oontroller installation, as opposed t,o Bafjoy -701 flod controller.

2 @ Esn Jose 23A4137 Rey,toC.provide any changes in Level 1, para. (a), ah. 42 of the as shown on sheet of this FDDR.

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408 925 5946 Oct 29,86 18: 41 F.04 TEL No. '

' GENERAL ELECTRIC P'DDR kT4-1680 ' RSV.'l2 JHERT 3ep 1r E*u'187 m4 O mociussuanov 8ENIRAL$ILtOfRIC mov 0

$UllN8SS OPanAtlows -_

Criteria Lavei_1 t h. .,.a1 to . sreater tue loot m a) ne aver. pu., 411.s4, . fle. A.

rated value afturggrseconds have elapsed frog)automatto $at$*1stion at any %n reacter pressure *>etween 200 pets (10.S kg/en' and rated.. .

b) The HPCI turbine shall act trip er isolate during aute er assusi start test s . '

14 vel 2 e) la order to provide su everspeed sad toelaties trip evoidatee'enrats, the transiest start first peak shall set some eleser than 135 (of tated speed) to the everspeed trip. sad subsequest speed peake shall met he grgater

than 31 above the rated turbine speed.

b) De speed and flow sentrol 1eeps eball be adjmated se ther the decay rati+ ,

of any EPCI systes relate ( vari.able is set greater than. 0.18.

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6 4) The delta-P evitch for the NFCI steam supply line high flow isolation trat shall be es11brated to actuate at between 2725 and 3003 steam s flow, and the value entered in the Plaat Technical Specifications.

e) MPCI discharge through the Core Spray $ystes must be 1300 1 500 CFM.

f)

NPCI dischstge through the Peedwater systes suas be 3100 g 500 0FM.

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NUCLEAR PRODUCTS AND ENGINEERING SERVICES October 28, 1996 - h.

g if tes AE Rogers To J. K. Sawabe subjects Hope Creek HPCI Startup Time References FDDR KT1-1691, Rev. 1,"HPCI System Flow Controller",

g October 29, 1986.

y f This letter documents the saf ety evaluation of the referenced FDDR f or

'{- the Hope Creek HPCI system flow controller.

The Hope Creek FSAR ECCS

f. performance analysis is creasing the time to rated flow for HPCI to 35 seconds. The limiting unaffected by in-i single f ailure in the FSAR analysis for all break sizes and locations is the Channel A DC source which eliminates the HPCI plus one diesel generator, leaving 3 LPCI + 1 CS loop + ADB available. All single Os failures which leave HPCI available have at least as many low pressure ECC systems available (see FSAR Table 6.3-6) and thus will not become limiting if the HPCI startup delay is increased to 35 seconds.

The HPCI system is also used along with the RCIC system to provide high pressure makeup in the event of a loss .of f eedwater transient.

Both the HPCI and RCIC are designed to individually maintain the vennel water level above Level 1 during this event. Since the HPCI flow capacity is larger than RCIC by nearly an order of magnitude, increasing the time to rated flow for HPCI to 35 seconds will not affect its ability to turn the water level around before Level 1 is reached.

Theref ore, it is concluded that increasing the time to rated flow for the Hope Creek HPCI system to 35 seconds is acceptable from a safety analysi s viewpoint.

ver m ed.

A? u D. A. Hamon, Technical Leader

%.9.QlMc J. R. Pallette, Technical Leader Plant Performance Engineering Plant Performance Engineering M/C 763, Ext. 54593 M/C 763, Ext. 56949 0

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k, 3.2 2 The fe!!awins paragraph shall be added to Paragraph 4.9.19 in epecifteeties. p he '

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( < Vaevue pump discharge line, frem the spetream s'ide of the shelk waive te ,

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! 3.2.3 choose the first sentence of Paragraph 1.1.1 in the bvee specificeth to .,.

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1.eee-of-Coelaat Accident (LOCA). Anticipated treatient Withest germ (43118). I or reeeter toelation and failure of the geacter Core teetaties Ces1(as (acic) system.

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3.t.4 The following paragraph empersedes Paragraph 4.11.5 la the beoe specification. -

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' The initiating and eteelag logic for the RPCI injectica velge to the WW line P shall be the same as the IPCI injection velve to the CN lino. Ze addition.

""~g 4, operatten of the injectica valve to the foodwater line at low reactee pressure N

, shall ebu t.he la eccordance with Paragraph 3.1.14 of the design specificattee date 3

3 3.2.5 W fe11 ewing paragr. A evpersedes Paragraph 4.9.6 ha the base i

specification. " .

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e N etarting time for the RPCI nyeten receipt of en actuation sissal to .' ' ' ' '

delivering design flow shall be within seconde.

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10ENERAL ELECTRIC TEL No- 408 925__5946 Oct 29.86,1@: 49, P,. 0g

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1 ,M a) The average pump distcharge) flow most be equal te or greater" tam 100s M rated value afterggseconds have elapsed frog auteestic 1a1 ties at any A.

reactor pressure .estween 200 pois (10.3 kg/ca*) and rated. m b) The HPCI turbine sh;di mot trip or isolate during aute er assusi acert teste.

'd,ll1 a) la order to provide sa overspeed sad isolattee trip ave &deseeJhdrata, the traastet., start first peak shall set some eleset than 3.55 (of gated speed)

- to the overspeed trip, and subsequent speed , seeks shall sat be gegator than 31 above the rated turbine speed.

b) The speed and flow control loops shall be adjusted so that the decay ratt+

of any MPCI syetes relatet t vertable is not greater thas 0.23.

c) The turbine gland seal eendenser system shall be aspeble ti preventing steen leakase to the staesphere.

d) The delta-P switch for the NPCS steam supply ilme high flow taolation trip shall be calibrated to actuate at between 2722 and 3005 staan flow, and the value estered in the Plaat Techsteal Specificettene. s e) MFCI dischstge through the Core Spray Systes must be 2500 1 500 GPN.

f) MPCI dietharge through the Feedwater System must be 3100 g 300 GPM.

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