ML20045E895

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Summary of 930519 Meeting W/Util Re Discussion of LPCI TS & Ielp Proposed TS Change for Plant.List of Meeting Participants Encl
ML20045E895
Person / Time
Site: Duane Arnold 
Issue date: 06/22/1993
From: Pulsifer R
Office of Nuclear Reactor Regulation
To:
Office of Nuclear Reactor Regulation
References
NUDOCS 9307060191
Download: ML20045E895 (26)
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Text

{{#Wiki_filter:s 4 pa%y j f .A i E UNITED STATES i( -) ! NUCLEAR REGULATORY COMMISSION ,/ s wAsmNGTON. D.C. M555-Md g June 22, 1993 1 Docket No. 50-331 LICENSEE: Iowa Electric Light and Power Company (IELP) FACILITY: Duane Arnold Energy Center (DAEC)

SUBJECT:

SUMMARY

OF MEETING HELD MAY-19, 1993 WITH IELP TO DISCUSS LPCI TECHNICAL SPECIFICATIONS On May 19, 1993, members of the NRC met with representatives of IELP to discuss the low pressure coolant injection (LPCI) Technical Specifications (TS) and the licensee's proposed TS change for DAEC. The meeting participants are listed in Enclosure I. The DAEC representatives presented an agenda for the discussions which is attached as Enclosure 2. The discussions in the meeting were to assure understanding of the LPCI design bases at DAEC and how TS Sections 3.5.A.5 and 6 can be clarified to reflect the intent. General discussions were held over the DAEC low pressure Emergency Core Coolant System (ECCS) design which includes three subsystems, two core spray subsystems and one LPCI subsystem, and that 2 of the 3 subsystems are required to meet the ECCS design bases. IELP discussed the loss of LPCI function versus the loss of the low pressure ECCS function. The NRC noted that several plants similar to DAEC have revised the loop select logic and now have a divisionalized Residual Heat Removal system, however, IELP said it would not be cost beneficial for DAEC. It was also noted that the proposed bases pages should be changed to more closely represent the low pressure ECCS 2 out of 3 subsystem requirement. There was general agreement on the proposed TS language to more clearly represent the LPCI Limiting Conditions for Operation requirements and on the proposed TS interpretation that will be in effect at DAEC until the proposed t i E n , m.... 290013 9307060191 930622 DR ADOCK 0500 1 gi [

LPCI TS changes are resolved by the staff. IELP agreed to delay issuance of the TS interpretation until the amendment request has been docketed. 6/ Robert M. Pulsifer, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Enclosures:

4 1. Attendance List 2. Meeting Agenda cc w/ enclosures: See next page DISTRIBUTION l Docket File (50-331) ACRS l l NRC & Local PDRs GGrant (17-G-21) l PDI11-3 Reading EGreenman, Rill TMurley/FMiraglia JPartlow JRoe JZwolinski j JHannon MRushbrook RPulsifer 0GC EJordan (MNBB 3701) RFraham (8-E-23) CSchulten (ll-E-22) ] 0FFICE PDIll-3:LA:DRPW PDIII-3:PM:MPA PDIIIND:DRPW MRu[6780k RPulsiferk/Y JHannon NAME 1 j \\. [h %[93 [f /21/93 [/ h93 DATE 0FFICIAL RECORD DOCUMENT NAME: G:\\DUANEARN\\LPCI. HTS ..r t 4 O

l l l ' LPCI TS changes are resolved by the staff. IELP agreed to delay issuance of the TS interpretation until the amendment request has been docketed. t obert M. Pulsi r, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Enclosures:

1. Attendance List 2. Meeting Agenda cc w/ enclosures: See next page 4

LPCI TS changes are resolved by the staff. IELP agreed to delay issuance of the TS interpretation until the amendment request has been docketed. Robert M. Pulsifer, Project Manager Project Directorate III-3 Division of Reactor Projects III/IV/V Office of Nuclear Reactor Regulation

Enclosures:

1. Attendance List 2. Meeting Agenda cc w/ enclosures: See next page DISTRIBUTION Docket File (50-331) ACRS NRC & Local PDRs GGrant (17-G-21) PDIII-3 Reading EGreenman, RIII TMurley/FMiraglia JPartlow JRoe JZwolinski JHannon MRushbrook RPulsifer OGC EJordan (MNBB 3701) RFraham (8-E-23) CSchulten (11-E-22) e) 0FFICE PDIII-3:LA:DRPW PDIII-3:PM:pBPM PDIIIND:DRPW NAME MRu hk80k RPulsifer[k/M JHannon \\ [x %[93 [f /21/93 h h 93 DATE 0FFICIAL RECORD 1 l DOCUMENT NAME: G:\\DUANEARN\\LPCI.MTS ) i i l 4

~ ~ j Duane Arnold Energy Center i Iowa Electric Light and Power Company cc: Mr. Lee Liu Chairman of the Board and Chief Executive Officer Iowa Electric Light and Power Company Post Office Box 351 Cedar Rapids, Iowa 52406 Jack Newman, Esquire l Kathleen H. Shea, Esquire Newman and Holtzinger 1615 L Street, N.W. Washington, D.C. 20036 J Chairman, Linn County Board of Supervisors Cedar Rapids, Iowa 52406 lowa Electric Light and Power Company ATTN: David L. Wilson Plant Superintendent, Nuclear 3277 DAEC Road Palo, Iowa 52324 Mr. John F. Franz, Jr. Vice President, Nuclear Duane Arnold Energy Center 3277 DAEC Road Palo, Iowa 52324 Mr. Keith Young l Manager, Nuclear Licensing 1 Duane Arnold Energy Center 3277 DAEC Road Palo, Iowa 52324 U.S. Nuclear Regulatory Commission Resident Inspector's Office Rural Route #1 Palo, Iowa 52324 Regional Administrator, Region III U.S. Nuclear Regulatory Commission 799 Roosevelt Road i Glen Ellyn, Illinois 60137 Mr. Stephen N. Brown Utilities Division Iowa Department of Commerce Lucas Office Building, 5th Floor Des Moines, Iowa 50319 l 1 .q ]

DAEC/NRR LPCI MEETING ATTENDANCE i May 19, 1993 HAME TITLE ORGANIZATION TELEPHONE Robert Pulsifer Project Manager NRR/PD III-3 (301) 504-3016 Jim Kinsey Licensing Supervisor IELP (319) 851-7177 Greg Whittier Systems Engineer IELP (319) 851-7496 Tony Browning Princ. Lic. Engineer IELP (319) 851-7750 Ron Frahm Sr. Reactor Engineer NRR/SRX8 (301) 504-2866 Carl Schulten Sr. Reactor Engineer NRR/0TSB (301) 504-1192 3 I f i 1 i I I

a i LPCI Tech Spec Meeting Agenda l Introduction Low Pressure Coolant Injection (LPCI) Subsystem ~ Description LPCI Subsystem Design Basis Accident Analysis Summary LPCI Tech Spec History Need for an Amendment Request Proposed LPCI Tech Spec Amendment L e I b 'f i P r f a ? I I - a- - - -. m L me

e \\ WN f - Valve 5*jnalled Se CLofE Yalve y nake) fo ope # 1ees 11 g,,.3.iif,37,2 s , (r,LC) r, (LC) 2001 2000 1, 1902 1903 1C03 VM Wl m s m, / V' V' y I l ICO3 31 1CD3 Jet y y-20-82 ZS Pumps N, Y-18-148 'g k lEm 's g 2 a A g u - Surge taas k> R.ase Pump i 13 1 P Pump 0y ou ic= 1c0 c~ c,. _. 3 19.. 1,,, ..,m, x oz 2005 2004 1933 1932 j y,,,,, Im ,,/ t ) \\ Sprar Yr -v,

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\\ i,, r, n.a., mm m h 5 1934 2007 / 3 1C04 2030 W 0)194o gg209 zii.eg pg =0n -8 "" __1. 19e gi rm 2009 1935 o 1913 b k 3012 i i ~ A 3 3 lM EM 20n 1912 e_ u RHR Serewe p 7 RMR Sernce 1.:8 a M ___ q W8 W^ Ifn 3,3 Figure 2. l PCI Mode of RHR System C-1: Residual liest Removal System 29

FOR INFOT:MATION alca tov-wu DRYUELL REACTOR A PRESSURE WATER LEVEL U,p 8:p ;, U Rev. 5 - I1/05/92 '{ 2 psig j 119.5" I 1-of-2-:vi:e l l1-of-2-tvice l l ARI BOTH i x RIC:RCT.u ! ION PUMPS RLNNING7 OF >2 psid F across pu=ps i YES ( )

0 1-of-2-tvice l

Logit Seal In fl points reset with j ' 1/2 secon: T.D. I Reser Loop Selection / l Break Detection Reset F REACTOR Pus hbut ton

  • PPlSSURE Time for Recire.

Y.,., < 900 psig , Pump Coastdown/ ,_{- } 1-of-2-twice fRecire. Loop i p, g Stabilization $/ '9eal In cannet 2 second *.D. l te reset after 450 psig inter-V RECIRCULATION icck unless 2s IS p, -7,>1 p s id ? L SYSTEM RISER and 119.5" are 1-of-2-tv1:e p ops clear (Figure 12) l h l 1/2 second T.O b ,./ f YES l f No l l If no Racir:. loop break, selects "B" loco -l h b b b b a FCLOSE A". TERMISSIVE REACTOR REACTOR 7ERMISSIVE $ 51 "B" i RECIRC. FOR PRESSURI PRES $URI TOR RICIRC. 10 MINUTES [<450 psig <450 psig 10 MINUTIS LOOP LOOP DISCRARGE/ DISCHARGE / g_og_2.gyge. l 1 og.2-twice d DISCHARGE DISCRARGE ,s BY? ASS h p g ngo gg BTPASS a v VALVES LPCI LOOP $ M LOOP "B"+ LOOP LPCI LOOP VALVES ..A" Recir,** INJECTION INJECTION ENJECTION INJECTION "B" Recir VALVES VALVES VALVES 7ALVES Pu=p Trips Pump Trips if not (Tigures ELOCK OPEN BLOCK OPEN (Tigures if not la & 15) 00T30ARD OUTBOARD 14 & 15) tripped by: tripped by: i (THROTTLE) (TIRDTTLE)

1) Leoo seect Loom yAI,vE VALVE
1) Leee 8* Wet Loom TOR 5 TOR 5 i

MINUTES MINUTIS (Tigures (Figures la & 15) 14 & 15) W W4?}k'CWQ Y M.ruTecT. CON ll( t,ccf S GLECTEb Figure 11. LPCI Loop Selection Logic Pbg 1AI TCcT'.Iod C-1: Residual lIcat Removal System 33

f i _i o uo oo m" TO 4.16 KV Bl)S l A3 (DIV. 1) TO 4.16 KV BUS l A4 (DIV. II) ] t o o s i e 480 V LC 183 480 V LC 184 e i e i e e i e i e 43/H l e Oi Oi l52-303 1 52-40 3 -- - - _ - - - - _ - - 4 p-____----g--,i e os os i i e i e - " 43/E T 43/E l i ---_---_--j[-------y l L-_-- e i l 43/E

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i e w i 480 V HCC 1834 480 V KC 1844 e i e i 8 i l i a i l E l l xi l 03 03 i %M C l e l52-3402 152-4402 i i g nls e os' oe i o-e .~ a ; l i Z Wl -_ E 43/E -ri e O O ,e Oi 52-4 401 - - - - - - - - - - - j s'" i Oi l 52-3401. - - -- -- _ - INTERLOCK j pg c t---_----_ --4---_ og DC TRIP ADDED oj DC TRIP ADDED i].e 32 43/N 480 V K C IB34A 480 V HCC 1844A _, -i ~;0 O C lllii;; aM C 22 O t-"" Z 4; DAIC SVING BUS F0ll0WlHG N0lfICATIONS

j MAJOR DIFFERENCES BETWEEN LOOP SELECT AND DIVISIONALIZED RHR All LPCI Flow Injected into ONE Recirc LOOP = = Crosstie Valve OPEN Recirc Discharge & Bypass Valves on " Broken" Loop = do NOT get a Closed Signal Swing Bus powers ALL LPCI Inject Valves and ALL = Recirc Discharge / Discharge Bypass Valves from one Division /EDG . ~ ;;q GQR y c r.,c. c u ;,.......,. 1 ~ c.. :. ~-.y 5 ) e 'E ?

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,= q. =_.=. i_af-pq?!FM.;;fiff)[M .. r.'!~ L . I.. - - ix..J '~~. ' :. HIGH O O PRESSURE LADS 1 HPCI Q Q CORE COOLING LOW i LPCI lV' PRESSURE l CORE s COOLING

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  1. Pumps Available LPCI Function Cont. Cooling Function (14,400 gpm Required) 4 Pumps OK OK (19,200 gpm) l 3 Pumps OK (Lose Redundancy)

OK (14,400 gpm) \\ 2 Pumps Not OK OK (%00 gpm) I 1 Pump Not OK OK (Lose Redundancy) (4800 gpm) i 0 Pumps Not OK Not OK .............___.......~....................... I Inject Viv Failure Not OK OK Swing Bus Failurs Not OK OK Cross Tie Valve Closed Not OK OK I l i

i i NkO 3 UFSAR/0AEC-1 ' ' /, -[l ~- r.: 6.3 EMERGENCY CORE COOLING SYSTEMS Y'* 6.3.1 DESIGN BASES AND

SUMMARY

DESCRIPTION This section provides the design bases for the emergency core cooling j systems (ECCS), formerly the core standby cooling systems (CSCS), and a i summary description of these systems as an introduction to the more detailed design descriptions provided in Section 6.3.2 and to the performance analysis provided in Section 6.3.3. 6.3.1.1 Desion Bases 6.3.1.1.1 Performance and Functional Requirements j The ECCS is designed to provide protection against the postulated l 9 LOCA caused by ruptures in the primary system piping. The functional I reauirements (i.e., coolant delivery rates) specified in detail in Table l 6.3-1 are such that the system performance under all LOCA conditions j postulated in the design satisfies the requirements of 10 CFR 50.46. I 3 These requirements, the most important of which is that the post-LOCA peak 1 cladding temperature (PCT) be limited to 2200 F, are summarized in Section l 6.3.3. In addition, the ECCS is designed to provide the following: I

1. Protection is provided for any primary system line break up to and including the double-ended break of the largest line.
2. Two independent phenomenological cooling methods (flooding and spraying) are provided to cool the core.
3. One high pressure cooling system is provided, which is capable of maintaining the water level above the top of the core and preventing automatic depressurization system (ADS) actuation for small breaks.

4 Automatic actuation is provided such that no operator action is J required until 10 min af ter an accident, to allow for operator assessment and decision,

5. The ECCS is designed to satisfy all criteria specified in this section

) for any normal mode of reactor operation.

6. A sufficient water source and the necessary piping, pumps, and other hardware are provided so that the containment and reactor core can be flooded for possible core heat removal following a LOCA.

6.3.1.1.2 RELIABILITY REQUIREMENTS l l The following reliability recuirements apply:

1. The ECCS conforms to all applicable requirements for redundancy and separation.

l

2. To meet the above requirements, the ECCS network has built-in redundancy so that adequate cooling can be provided, even in the event of specified failures.

As a minimum, the following equipment makes up this system: 6.3-1 Revision 9 - 6/91

h UFSAR/DAEC-1 /*~ty*r's. dt -

a. One high-pressure coolant injection (HPCI) system.

hj.'.s.R., 2 c, '., ' ' n'. b. Two core spray (CS) systems. ..*r t,

c. One low-pressure coolant injection (LPCI) system.
d. One automatic depressurization system (A05).
3. The ' system is designed so that a single active component failure, including power buses, electrical and mechanical parts, cabinets, and wiring, cannot disable the automatic depressurization system.

i 4 If there is a break in a pipe that is not a part of the ECCS, no single component failure in the system prevents automatic initiation and successful operation of less than one of the following combinations of ECCS equipment: a. Two LPCI pumps, one core spray loop, the automatic depressurization l 9 system, and the HPCI system (i.e., single diesel generator l failure). I 8 b. Two LPCI pumps, one core spray 1o00 and the automatic l depressurization system (i.e., Division II 125V battery failure). l-c. Four LPCI pumps, two core spray loops, and the automatic l .9 depressurization system (i.e., HPCI f ailure). I I d. Two core spray loops, the HPCI system, and the automatic i-9 depressurization system (i.e., LPCI injection valve failure). I t 5. If there is a break in a pipe that is a part of the ECCS, no single 1 9> component failure in the system prevents automatic initiation and i successful operation of less than one of the following combinations I of ECCS equipment: 1 a. Two LPCI pumps, the HPCI system, and the automatic depressurization i 8 system (core spray break with a concurrent diesel generator I failure). I b. Two LPCI pumps and the automatic depressurization system (core 1 9 spray break with a cencurrent Division II 125V battery' failure). I i

c. One core spray, HPCI system, and automatic depressurization system j

(core spray, LPCI injection valve f ailure).

d. Two LPCI pumos, one core spray loop, and automatic.depressurization 1 8:

system (HPCI break with a concurrent diesel generator failure or l HPCI break with a concurrent single 125V battery failure). .I

e. Two core spray loops and automatic depressurization system (HFCI break, LPCI injection valve failure).

These are the minimum ECCS combinations that result after assuming any failure (from item 4 above), and assuming that the ECCS line break disables the affected system, l I 6.3-2 Revision 9 - 6/91 j

yp,A UFSAR/DAEC-1 fj,j p r p#, df. in case the core spray line breaks. A manual valve, which is normall .f - N locked open, is provided downstream of the inside check valve. The valve. : ' shuts off the core spray system from the reactor during shutdown to permit maintenance of the upstream valves. The two pipes in the core spray system enter the reactor vessel through nezzles located 180 degrees apart. Each internal pipe then divides into a seeicircular header, with a downcomer at each end that turns through the shroud near the top. A semicircular sparger is attached to each of the four outlets to form two circles, one above the other and both essentially complete. Short elbow nozzles are spaced around the spargers to spray the water radially onto the tops of the fuel assemblies. Core spray piping upstream of the outboard shutoff valve is designed for the lower pressure and temperature of the core spray pump discharge. The outboard valve and piping downstream are designed for reactor vessel pressure and temperature. All piping and pump casings are designed in accordance with the criteria presented in Chapter 3. The RHR/ core spray fill pump maintains system piping filled with 6 water to prevent the potential for water hammer as discussed in Section l 5.4.7.2.1. l The core spray equipment, piping, and support structures are designed in accordance with Seismic Category I criteria to resist motion effected by the DBE at the installed location within the supporting building. For seismic analysis, the core spray system is assumed to be filled with water. Low water level in the reactor or high pressure in the drywell signals the automatic controls to energize the core spray pumps and place system valves in the sprry mode. When reactor pressure decreases, the core spray shutoff valves are signalled to open. Flow to the sparger begins when the pressure differential opens the inside check valve. Section 7.3.1.1.2 gives further details and evaluation. 6.3.2.2.4 Low-Pressure Coolant Injection The LPCI system is an operating mode of the RHR system. The LPCI systa is automatically actuated by low water level in the reactor and/or high pressure in the drywell. It uses four motor-driven RHR pumps to draw suction ' rom the suppression pool and inject cooling water into the reactor tora. The LPCI system, like the core spray system, is designed to provide cooling to the reactor core only when the reactor vessel pressure is low, as is the case for large LOCA break sizes. However, when the LPCI system operates in conjunct ha with the automatic depressurization system, the effective core-cooling capWlity of the LPCI system is extended to all break sizes because the automatic depressurization system rapidly reduces the reactor vessel pressure to the LPCI operating range. Figure 6.3-3 is a schematic process diagram of low pressure coolant injection. LPCI operation is based on using three of the four ac motor-driven centrifugal pumps that take water from the suppression pool 6.3-14 Revision 8 - 6/90 l

[ f,..p. p. UFSAR/DAEC-1 s and pump it into onelof the two recirculation loops. The water enters the reactor through jet pumps and restores the water level in the reactor vessel. Figure 7.3-13, Sheets 1 through 3, is the flow control diagram for the RHR system including the LPCI system. The RHR/ core spray fill pump maintains system piping filled with ( 6 water to prevent the potential for water hammer, as discussed in Section l 5.4.7.2.1. I The LPCI pumps receive power from the 4160-V ac emergency auxiliary buses. For each loop, the LPCI pump motors and associated automatic motor-operated valves receive ac power from different buses. LPCI pumps and piping equipment are described in detail in Section 5.4.7. Also described are other functions served by the same pumps if they are not needed for the LPCI function. Portions of the RHR system required for accident protection are designed in accordance with Seismic Category I criteria. 6.3.2.2.5 HPCI, Core Spray, and LPCI Pump Curves Curves showing head, horsepower, net positive suction head versus flow, and efficiency for the HPCI, Core Spray, and RHR (LPCI) pumps are presented as Figures 6.3-4, 6.3-5, and 6.3-6. Specific speed for each pump is also indicated in these figures. Pump runout conditions could occur in certain situations where the RHR (LPCI) pumps discharge to flow paths with too little system flow resistance. The operation of the RHR (LPCI) pumps under this condition could result in damage to the pumps from cavitation and/or motor overload. The DAEC is in the category of BWR/3 and BWR/4 plants with loop selection logic systems (LSLS). The following situations could potentially result in RHR (LPCI) pump runout conditions and a subsequent reduction or loss of long-term heat removal capability following a postulated LOCA for this category of plant: 1. Four LPCI pumps injecting into a broken recirculation loop from a single LSLS failure. 2. Four LPCI pumps injecting into both recirculation loops simultaneously, with one loop broken, from a single LSLS failure.

3. Operation with three pumps providing flow (one pump inoperable as allowed per the Technical Specifications) to the unbroken loop, with the single failure of a recirculation loop discharge valve to close.

An evaluation was performed on the DAEC RHR system to determine possible effects on long-term heat removal capabilities. With respect to the above potential RHR runout conditions, no other situations were found to be more severe than conditions 1 through 3 above. Resistance calculations were performed on the RHR recirculation piping network to determine the loop with the highest RHR pump runout potential. The following network configurations were evaluated with respect to their associated potential RHR runout conditions: 6.3-15 Revision 8 - 6/90 l

1 UFSAR/DAEC-I t \\ vessel. Primary containment pressure is monitored by four pressure h'. 'S switches that are mounted on instrument racks outside the drywell but 'J j '- I' inside the reactor building. Pipes that terminate in the reactor building allow the switches to sense pressures within the drywell interior. \\ System controls function to provide makeup water flow to the reactor vessel until the amount of water delivered to the reactor vessel is adequate. The HPCI system then automatically shuts down. Controls for j remote manual startup, operation, and shutdown are located in the main i control room. Once actuated to ensure proper functioning, RPV steam must power the HPCI turbine-driven pump. Instrumentation installed to detect steam flow is necessary to indicate steam flow status. 6.3.5.2 ADS Actuation Instrumentation i The automatic depressurization system is automatically actuated by ( 3 i signals from instrumentation monitoring reactor water level. Reactor l vessel low water level signals actuate a time-delay circuit. In addition t to the time-delay circuit, core spray or RHR pumps must be running to I initiate reactor vessel blowdown. The automatic depressurization system l can also be manually actuated from the main control room. Automatic I actuation can be prevented from the control room during the time-delay I by placing the ADS timer reset switches in the override position. I l 6.3.5.3 Core Sorav Actuation Instrumentation Automatic start of both pumps is initiated by the instrumentation signals generated by either reactor vessel low water level or drywell high pressure (one-out-of-two-twice logic for either signal). In addition, the core spray can be manually actuated from the main control room. 6.3.5.4 LPCI Actuation Instrumentation Low pressure coolant injection is automatically actuated by the RPV low water level or high drywell pressure. In addition, low pressure coolant injection can be manually actuated from the main control room. The low pressure core cooling portion of the emergency core cooling systems consists of three subsystems: core spray A, core spray B, and low pressure cooeant injection. Therefore, it should be understood that the LPCI subsystem by itself is not required to meet all the requirements of IEEE 279, since it is backed up by the two core spray subsystems. To the extent practicable, the LPCI subsystem _has been designed to meet IEEE 279. The loop selection sensing instrumentation for break detection and valve selection is arranged so that the failure of a single device or circuit to function on demand will not prevent the correct selection of the loop for injection. The control system reliability is compatible with, and more reliable than, the controlled equipment (injection valve). Those single failures that could cause improper loop selection (i.e., selected short circuits that pick up specific relays) will not disable the core spray function. It is concluded, therefore, that the failure of the loop selection scheme 6.3-30 Revision 8 - 6/90 l

DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT Ltm Freouency c. Motor Once/3 months i Operated

  • ')

Valve ,;.L.. N [. Operability -{- Fm,' 9.g J " i 4 ,s.n' V d. eump rio-Once<3 months v Rate Three LPCI pumps shall deliver 14,400 gpm against a system head corresponding to a vessel pressure of 20 psig based on individual pump tests. 4. From and after the date that 4. When it is determined that one one of the RHR (LPCI) pumps is of the RHR (LPCI) pumps is made or found to be inoperable inoperable at a time when it is n d bon required to be OPERABLE, the eac o permissible only during the remaining active components of succeeding thirty days the LPCI subsystem, the 'E '# r a h ir active components of the '1 both core spray subsystems subsystem, the containment shall be demonstrated to be e cooling subsystem, and all OPERABLE immediately and the active components of both core OPERABLE LPCI pump daily there-spray subsystems and the after. diesel generators are OPERABLE. 5. From and after the date that 5. When it is determined that the two RHR pumps (LPCI mode) are LPCI subsystem is inoperable, made or found to be inoperable both core spray subsystems and for any reason, continued reactor operation is the containment spray subsystem permissible only during the shall be demonstrated to be succeeding 7 days unless at OPERABLE immediately and daily least one of the inoperable thereafter' pum;,s is sooner made OPERABLE, provided that during such 7 days all active components of. both core spray subsystems, the containment spray subsystem and the diesel-generators required for operation of such components are OPERABLE. Amendment No. 160 3.5-3 06/89

DAEC-I LIMITING CONDITION FOR OPERATION SURVEILLANCE. REQUIREMENT t 3. The LPCI Subsystem shall be 3. LPCI Subsystem Testing sha#1 be j OPERABLE whenever irradiated fuel 85 follows: is in the reactor vessel, and Item freauency l prior to reactor startup from a COLD CONDITION, except as a. Simulated Annual specified in 3.5.A.4, 3.5.A.5 and Automatic 3.5.G.3 below. Actuation Test b. Pump Once/3 months Operability c. Motor Operated Once/3 months Valve Operability d. Pump Flow Once/3 months Rate inree LPCI pumps shall deliver 14,400 gpm against a ,s n ry.j system head .. ? c :" corresponding b ^ A u -. ~ ~ ~,,, : v to a vessel r^S

  • i

~. r' pressure of u *~

  • 20 psig based on individual pump tests, e.

Once per shift visually inspect and verify that RHR valve panel lights and instrumentation are functioning normally. 4. With one RHR (LPCI) pump inoperable, provided the remaining RHR (LPCI) active components, both l Core Spray subsystems, the l l containment spray subsystem, and I l the diesel generators are verified l to be OPERABLE, restore the l 1 inoperable RHR (LPCI) pump to l l 1 OPERABLE status within 30 days. 5. With two RHR (LPCI) pumps inoperable, providing both Core Spray subsystems, the containment spray subsystem, and the diesel generators are verified to be l OPERABLE, restore at least one RHR l (LPCI) pump to OPERABLE status t within 7 days. [ l l Amendment No. 174 3.5-3 09/91 l L

4 DAEC-1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT 6. Otherwise, be in at least HOT SHUTDOWN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. B. Containment Sorav Coolina B. Containment Soray Coolina Capability Capability i 1. The suppression pool and drywell Surveillance of the containment i spray modes of the residual heat removal (RHR) system shall be spray loops shall be performed as OPERABLE with two independent follows: i loops each when the reactor water temperature is reater than 212*F 1. During each five year period, an {]t as speci ied in 3.5.B.2 and air test shall be performed on the 3 drywell and suppression pool spray 2. With one suppression pool spray headers and nozzles. loop and/or one drywell spray loop inoperable, restore the inoperable loop to OPERABLE status within 30 I days or be in at least HOT SHUTDOWN I within the next 12 hours and in l COLD SHUTDOWN within the following i 24 hours. A t.n 3. With both suppression pool spray V)f y loops and/or both drywell spray e6 loops inoperable, restore at least -P. ?%. one loop to OPERABLE status within [/,$.J 6%'e 8 hours or be in HOT SHUTDOWN within the next 12 hours and in ' + COLD SHUTDOWN within the following

  • V 1

24 hours. ' 'L g l Amendment No. 174 3.5-4 09/91

i DAEC-1 1 LIMITING CONDITION FOR OPERATION SURVEILLANCE REQUIREMENT ! 3. The LPCI Subsystem shall be 3. LPCI Subsystem Testing shall be i OPERABLE whenever irradiated fuel as follows: is in the reactor vessel, and Io Fr e my l prior to reactor startup from a COLD CONDITION, except as a. Simulated Annual specified in 3.5.A.4, 3.5.A.5 and Automatic 3.5.G.3 below. Actuation Test b. Pump Once/3 months Operability c. Motor Operated Once/3 months Valve Operability d. Pump Flow Once/3 months Rate ~ ' Three LPCI pumps shall deliver 14,400 gpm t, f.--C ,.. against a system head \\y' corresponding to a vessel pressure of 20 psig based on individual pump tests. e. Once per shift visually inspect and verify that RHR valve panel lights and instrumentation are functioning normally. l 4. With one RHR (LPCI) pump l inoperable, provided the remaining RHR (LPCI) active components, both l Core Spray subsystems, the l containment spray subsystem, and l l the diesel generators are verified ~ ~ ~ ' l to be OPERABLE, restore the N U C 5d5f5 h

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l l inoperable RHR (LPCI) pump to b be Loydle. T.e my rer. sod l 1 OPERABLE status within 30 days. - - -d l N, m g, 3, y., _ y mG ~'~ l_ N, 5. With - _--._v r.c?p3bl; JprovHling both Core ga UCJI. =,My54% Spray sub~s'ystems, the containment spray subsystem, and the diesel generators are verif d to be l store Ucq_r;,""1/ p g lfC1 sd + b q k N b ed 1 m y_,_j p o OPL ABLE status mya 1 g. g g g 4 ( I wi nin 7 days. . a % M 2. S L IA u Ceel y 4 , Mle <+ - m JI Owh.orlocJ We uc/ n-hel 4 .J a. Amendment No. 174 3.5-3 91 l

4 DAEC-1 I' "'- 3.5 BASES ~ j !X '. ~ - ~%....,_.,"'~- A. Core Spray and LPCI Subsystems This specification assures that adequate emeroency cooling capability is available whenever irradiated fuel is in the reactor vessel. i l Based on the loss-of-coolant accident (LOCA) evaluation models described in General Electric Topical Report NE00-20566 (Ref. 2), the results of the LOCA i analysis given in Reference 3 and Subsection 6.3 of the Updated FSAR and in accordance with the acceptance criteria of 10CFR50.46, any of the followinglcomkr.d,m E' I nt cooling to the core to dissipate the cooling systems provides energy associated with the oss-of-coolant acciden t; '4-it cc'culated ] feel : lad tcmpcrcturc to ic;; ther. 2200*T to essure toet cccc gcieAd -rc in; nt ct, :nd t0 'i-it clad etal a atcr rca;ticr to less ther, 1%; i / n... -,. -. - ['InserI A] The limitino conditions of operation in Specification 3.5. A.1 through 3.5. A.6 specify tne combinations of operable subsystems to assure the availability of the minimum cooling systems noted above. 04/83 3.5-14 Amendment No. 88

DAEC-1 Core spray distribution has been shown, in full-scale tests of systems similar in design to that of DAEC to exceed the minimum requirements. In addition, cooling, ef fectiveness has been demonstrated at less than half the rated flow in simulated fuel assemblies with heater rods to duplicate the decay heat characteristics of irradiatec fuel. The accident analysis is additionally conservative in that no credit is taker for spray coolant entering the reactor before the internal pressure has f allen to 151 psig. 54>syslm The LPCI subsystem is designed to provide emerge cy cooling to the core by flooding E in the event of -of-coolant accident. This :yst= functions in combination witt thecorespraysyste(.oprventexcessivefuelcladtemperature. The LpCl subsyster and the core spra subsystemf rovide adequate cooling for break areas of approximatel.s 0.2 square feet up to and including the double-ended recirculation line break withou-assistance from the high pressure emergency core cooling subsystems. l l The allowable repair times are established so that the average risk rate for repair l' would be no greater than the basic risk rate. The method and concept are described l. in Reference 1. j.,; i Using the results developed in this reference, the repair period is found to l f };-[," be 1/2 the test interval. This assumes that the core spray subsystems and LPCI L constitute a 1 out of 3 system; however, the combined effect of any of the two subsystems to limit excessive clad temperatures must also be considered. i The surveillance requirements provide adequate assurance that the Core Spray I subsystems and the LPCI subsystem will be operable when required. h5cdk3 t M c.b3 ShouldthelossofonefLPCI)pumpoccur, a nearly full complem t f core and containment spray equi ment is available. TheremainingthreeLPCl)pumpsand spray subsyste will perform the core cooling function. Because of the availability of the majority of the core cooling equipment, which will be verified to be operable, a thirty da ir period is justified. If the LPCI l P. R subsystem is not available, at least 2 LfCJ pump,, ust be available to fulfill the containment spray unction. The 7 day repair period is set on this basis. l A ) for k less.C %e. Kmendm I No. 174 3.5-15 N ^ 1'_h_

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09/91

-.3 l ~ Insert A: f

a. Two LPCI pumps, one Core Spray loop, the Automatic Depressurization system, and the HPCI system (11 single Diesel-Generator failure),
b. Two LPCI pumps, one Core Spray loop, the Automatic Depressur.zation system GA, Division !! 125V battery failure).
c. Four LPCI pumps, two Core Spray loops, and the Automatic Depressurization system UA, HPCI failure).
d. Two Core Spray loops, the HPCI system, and the Automatic Depressurization system (!A, LPCI injection valve failure).

( [ rte B; In addition to the normal OPERABILITY requirements of having a bydraulic flowpath from the Suppression Pool to the Reactor Pressure Vesse! at the required head and flowrate (3 RHR pumps @ 14,400 gpm and 20 psid), the LPCI subsystem must also be capable of directing this flow into the appropriate Reactor Recirculation loop. Consequently, the LPCI

  • Loop Select logic" instrumentation (Specification 3.2.B) and LPCI
  • Swing Bus" (Buses IB33A and IB43A) must also be OPERABLE to support the LPCI function.

Irtsert C: In the course of a plant shutdown, during the transition from Startup/ Hot Standby to Cold Shutdown, the RHR system must be re-aligned from its standby-readiness mode (i.e., LPCI) to the Shutdown Cooling mode. While the system must be manually re-aligned from the Shutdows Cooling mode back to the LPCI mode upon demand, the LPCI mode may be considered OPERABLE for the purposes of satisfying Specification 3.5.A.5, as the system will be Operating (i.e., known to be OPERABLE); this re-alignment is a normal plant evolution: and, the nsk of a LOCA during this transition from the RHR cut-out pressure (135 psig) to Cold Shutdown is mimmal. i i I l

( l i ~ DRAFT ItCHNICAL SPECIFICATION INTERPRETATION MEMORANDUM T.S. SECTIOst 3.5.A T8IN NO. 93-01 TOPIC l LPCI Subsystem Operability What is the correct LCO time, L.,h, Allowed Outage Time ( AOT), for a loss of the entire LPCI function? INTERPRETATIO0t l As the design basis of the LPCI subsystem is to deliver at least 14,400 gpe to the vessel, any combination of inoperable equipnent which defeats this design basis (e.gi, loss of an injection valve, the swing bus, the cross-tie valve or the common Torus suction valve) is treated as a loss of the entire LPCI function in the accident analysis. Since having 2 RHR pumps inoperable defeats the ability to provide the required flowrate, TS 3.5. A.5 governs the complete loss of the LPCI function and results in a 7-day LCO provided both Core Spray subsystems, a containment Cooling subsystem (including 2 RER Pumps) and the Diesel Generators are OPERABLE. BASIS l The LPCI mode of the RHR system is one of three subsystems of the Low Pressure Core Cooling System, along with the two Core Spray subsystems (TS 3.5.A). The loss of any one of the three subsystems, h, loss of either Core Spray subsystem or the LPCI subsystem, is bounded within the accident analysis (Ref. UFSAR 6.3.1.1.2). With the loss of any one of these subsystems, the accident analysis is still satisfied, but the plant is no longer capable of coping with another single active failure within the ECCS. A 7-day LCO is provided for these situations (TS BASES 3.5.A). Additional supporting information can be found in References 1 and 2 below. REFERENCES l

1) R. Anderson to K.

Young, et.al., OH-93-0028, Feb. 19,1993.

2) 7. Browning to File, NG-93-2039, March 10,1993.

Prepand by Date Supervising Engineer - Systerns Engineering Approved by Date Assistant Plant Superintendent - c'perations Support Approved by Date Manager, Nuclear Licensing OC Concurrence Date NG-041G Rev 1 (102.16) {}}

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