ML20236G577
| ML20236G577 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 10/23/1987 |
| From: | Allen C COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| 3756K, TAC-59107, TAC-59108, NUDOCS 8711030095 | |
| Download: ML20236G577 (17) | |
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. N Commonwealth Edison
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f One First Natior,d Plazai Chicago, Illinois
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7 Eddress Reply to: Post Office Box 767 I
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- Up October' 23, 1987 y:
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i c,yp U.S. Nuclear Regulatorf Commission
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Attn: Document Control Desk 1,
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Subject:
.LaSalleCountyStationUnits))and'2.-
3 Implementation of 10 CFR 50-61f MW Rule.
b NRC Docket, Nos. 50--373 and 50-714
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TAC No.'SP107, 59108-9 1
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Reference (a):- Letter Lateid December 19,: 19F7 from A. Bournia to D.L. Farrar P
1 y(b): Letter dated April 6, 1987 from C.M.' Allen I
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( (c): Letter dated June 12,.1987 from D.R. Muller to D.L.'Farrar
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Dear Sir:
1 Attached is Commonwealth Edison's responfje for LaSalle'Countyl L
Station regarding compliance with the requirements of 10 CFR 50.62<
Anticipated. Transient without Scram. This information was originally requested in referer.ce (a). Reference.(b) transmitted a partial response.
This response ruplaces reference (b) in it's entirety.
h Attachment /1fdf.theresponsedescribestheproposedARIsystem design for LaSalle' County Station and has indicated that it satisfies.the requirements of l'O CFR 50.62'(c)(3). Attachment 2 describes the evaluation of the ATWS Recircular).on dump Trip (RPT) system dnd its conformance to the requirements ofcl0 CFR 50.62 (c)(5). Attachment Yrdescribes implementation of 10 CFR 50.62(c)(ts) requirement for the Standby Liqbtd Control System.
Included in that attachmentfis a response to the request for. additional:
I information transmitted in reference (c).
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If you have jany addf.tional questions regarding thirs matiter, please contact this office.
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AHAC M 1 IMPLEMENTATION OF 10 CFR 50.62 REQUIREMENTS FOR THE ALTERNATE ROD INJECTION (ARI) SYSTEM AT LASALLE COUNTY STATION - UNITS 1 AND 2 REQUIREMENT - 10 CFR 50.62(c)(3):
j Each boiling water reactor must have an ARI system that is diverse (from the reactor trip system) from sensor output to the final actuation device. The ARI system must.have redundant scram air header exhaust valves.
The ARI must be designed to perform.its function in a reliable manner and be independent (from the existing reactor trip system) from sensor output to the final actuation device.
LASALLE IMPLEMENTATION The ARI system design for LaSalle County Station (LSCS), Units 1 and 2, satisfies the requirements of 10 CFR 50.62(c)(3). The following text' correlates with the NRC SER Appendix A checklist enclosed with Attachment 1.
1.
ARI System Function Time The ARI system function will begin within 15 seconds and be completed within 25 seconds.
The results of the ARI transient analysis described in analysis report ARI System, Quad 1-83-007, indicates that the ARI function time objectives are met as follows:
I a) MSIV Closure (MSIV) l The ARI is initiated 15 seccnds after a recirculation pump trip on High Reactor Vessel pressure (1135 psig) and is completed within 3.67 j
seconds.
The total time from MSIV closure start to the completion of the ARI scram is 21.88 seconds.
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l b) Turbine Trip (TTWB)
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The ARI is initiated 15 seconds after a recirculation pump trip on High Reactor Vessel Pressure (1135 psig) and is completed within 3.67 seconds. The total time from turbine trip to completion of the ARI scram is 20.16 seconds, c) Inadvertent Opening of Safety Relief Valve (IORV)
The ARI is initiated manually when the suppression pool temperature reaches 110*F.
After a 15 second delay the ARI scram begins and is completed within 3.9 seconds. The total time between the manual initiation and the PRI scram completion is 18.90 seconds.
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2.
Safety-Related Requirements The ATWS Rule does not require the ARI system to be safety grade; however, i
the implementation of ARI must be such that the existing protection system, i
continues to meet all applicable safety-related criteria. Specifically, j
qualified isolators should be used-for ARI system interfaces with safety systems.
i The ARI design at LSCS, Units 1 and 2 is safety-related and interfaces with two electrical divisions, ESP Division.1 and ESF Division 2.
The i
equipment at the interfaces is qualified to IEEE Standards 323 and 344.
isolation devices are. required at the ARI/ Safety-System interfaces.
j Divisional physical separation and independence is maintained. No Interfaces between the ARI and non-safety related readouts are provided by Class lE relays where the safety-related relay coil is divisionally
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associated and the isolated relay contacts are used for the non-safety related. readouts.
3.
Redundancy The ATWS rule requires that the ARI system must have redundant scram air j
header exhaust valves and perform a function redundant to the backup scram system. ARI self-redundancy is not required.
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The LaSalle design employs redundant dual-coil solenoid-operated valves to I
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vent the scram air header, to block the instrument air supply line to the pilot scram valves, and to exhaust the air header to the scram discharge volume vent and drain valves permitting those valves to close. The 1
one-out-of-two taken twice logic of either ESF division 1 or 2 will initiate the ARI.
4.
Diversity from Existing Reactor Trip System (RTS)
The ATWS Rule requires that the ARI system equipment be-diverse from the existing reactor trip system to the extent reasonable and practicable to minimize the potential for common cause failures from the sensors output te, and including, the components used to vent the scram air header.
The LaSalle specific design satisfies these requirements as follows:
a) Energize-to-Function The ARI system valves have separate, energize-to-open, dual coil, solenoid valves.
b) DC Powered Valves The solenoid valves are powered from the respective divisionally associated 125 Vdc ESF buses (1 or 2)
m c)l Instrument-Channel Components
-The ARI channel components! employ" linear analog transmitters.and trip.
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' units which are ~ separate sensors =from -those:used for.the Reactor Protection System (RPS). Therefore,.the ARI conforms to the ATWS Rule equipment ~ diversity. requirements.
5.
Electric-Independence'from'the Existing RTS I
a) The'ARI design for LSCS is safety-related with power supplied from
'two electrical divisions; ESF division:1 and ESF divisionL2.--The~RTS at LSCS:is the Reactor Protection System (RPS), which is a 4-channel'-
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' electrical arrangement.which has' individual [ channel separation.- All
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four-RPS channelsfare. routed.in separate conduits and never routed l
l with ESF' division"1.:and 2-(ARI)..The separation criteria between RPS
'l subchannels and ESF divisions is'further. described in Chapter 8 of-i the LSCS FSAR..
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'b)l The ARI systemLfor LSCS is designed as'a safety-related system.
The-only interface with a non-safety related system is the. station-annunciator and' digital computer where independent' contacts of safety-related relays are used as inputs.
6.
The ARI system at LSCS-is designed'to-be physically separated.from the RTS (which is the RPS system at LSCS). The safety-related design of this
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l system with ESF separation criteria ensures that. physical separation is-maintained.
- 7. -N/A l
8.
N/A 9.
Safety-Related Power Supply.
I a) The power for the RTS (RPS) is 120 Vac from the RPS motor generator l
sets. The power for ARI is from 125 Vdc safety-related buses.
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b) The 125 Vde' buses remain energized during loss of offsite power events.
10.
Testability at Power ~
The ATWS Rule guidance stated that the ARI system should be t'estable at ~
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. ower. The LaSalle design conforms.to the Rule as follows:
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!- a) Testability The ARI control logic can be tested during plant operation. Test
. switches are activated by the control room operator to~ prevent the opening of the ARI solenoid valves inadvertently. The divisionally.
separated channel trip logic requires one out of two taken twice (1/2 x 2) Reactor Vessel Level 2 "or (inclusive)" one out of two taken twice (1/2 x 2) Reactor Pressure Vessel Head Pressure-high to trip its divisionally. associated scram. valve coil. The scram air header will-vent on. receiving an actuation signal from either ESF division 1 (channels A "or" C) or division 2 (channels B "or" D) trip logic.
Therefore, the at-power testing of either division is permitted
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without disabling the ARI system.
b) Indication of Bypass Test switches and (white) indicating lights are provided for testing q
the ARI control logic. These lights provide test status' indication 1
in the control room at panel H13-P603.
A red indicating light is i
also located on the same channel to provide an indication that the tested channel logic will provide an initiating signal.
The conformance of the LaSalle ARI design to the requirements of Reg.
i Guide 1.47 is identical to that which was-accepted.by the NRC for the safety-related reactor t, p system, i.e., in accordance with the applicant /AE interpretation due to the promulgation of Reg. Guide 1.47 after the issuance of the construction permit for LSCS.
The bypass or deliberately induced inoperability of the ARI is indicated at the system level by manual initiation of the coil test switch located on panels H13-P603 as part of the administrative procedures.
Therefore, the ARI is in conformance to the requirements of L g.
Guide 1.47.
11.
Inadvertent Actuation The ATWS Rule guidance states that the inadvertent ARI actuation which challenges other safety systems should be minimized. The LaSalle ARI design conformance to the Rule as follows:
a) Scram Challenge
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The trip settings for.the reactor pressure vessel (RPV) head pressure (1135 psig) and the RPV level 2 water level (-50" with respect to-instrument zero) compared to the existing RPS trip settings,'as i
defined in the LaSalle Technical Specification Table 2.1.1-1, indicate a difference between reactor trip system and ARI system level and pressure setpoints of greater than or equal to 62.5" and less than or equal to 92 psig, respectively. Therefore, the ARI actuation setpoints will not challenge the scram.
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. b.
Coincidence Logic The coincidence actuation logic described previously in 10(a) is also intended toiprevent spurious actuation.
Single _ channel failures will not result in a coil actuation signal.
12.
Manual Initiation Two manual initiation switches are provided for each divisional control logic. These switches are located in close proximity to the existing RPS manual scram pushbuttons. Failure of an automatic ARI initiation cannot prevent manual initiation. In order to avoid an inadvertent manual trip of the ARI system, the two manual initiation switches in each divisional control logic must be activated to permit manual initiation of the ARI system.
13.
Information Readout The ARI design provides continuous readout of-level and pressure on the master trip units located on panels H13-P800 and H13-P801 in the auxiliary electric equipment room (AEER).
It also provides trip status, valve position, test status, inoperative, failure, and maintenance status at panel H13-P603 in the control room.
ARI system unique annunciators also provided on the main operator console in the control room for each ARI channel to indicate that:
a.
The ARI system has been initiated, and b.
The ARI system has timed out after two minutes.
Indicators are also provided for input trip signals to the ARI and output protective action signals to the scram valve coils. Red lights indicate a logic trip and amber lights indicate manual reset permissive for the trip logic after a two minute seal-in period.
14.
Completion of Protection Action In the LaSalle ARI design, once the ARI system is initiated, the solenoid operated valves will be energized to initiate a reactor scram. The automatic and manual actuation signals to the ARI valves seal-in for two minutes to assure that all control rods have time to fully insert.
A summary of the LaSalle Spccific ARI system features identified in the -
letter from Anthony Bournia (NRC) to Dennis L. Farrar (CECO) dated December 19, 1986 for which the above information has been requested is given in the enclosed completed SER Appendix A Checklist with additional comments. The additional comments provide traceability to LaSalle specific documentation which support the above information.
.APPENDIXA CHECKLTST FOR PLANT SPECIFIC' REVIEW'OF ALTERNA2E ROD INJECTION SYSTEM (ARI)
ARI system' function time Rod injection motion will begin with 15 seconds and be completed within 25 seconds from ARI initiation.
Yes 2.
Safety-related requirements a) Class lE isolators are used to interface with safety-
.Yes, Note 1 related systems l
b) Class lE isolators are powered from a Class lE source Yes, Note 1 c) Isolator qualification docum1nts are available for staff Yes, Note 1 audit.
3.
Redundancy The ARI system performs a function redundant to the backup scram system Yes 4.
Diversity from existing RTS a) ARI system is energize-to-function Yes b) ARI system uses DC powered valves Yes c) Instrument channel components. (excluding sensors but including all signal conditioning and isolation devices) are diverse from the existing RTS components.
Yes i
5.
Electrical independence from the existing RTS a) ARI actuation logic separate from RTS logic Yes, Note 2 b) ARI circuits are isolated from safety-related circuits Yes, Note 3 i
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'T' i-Conformance'
- with ARI SER 6~. ~ Physical. separation'from the existing RTS (a) ' ARI: system is ' physically: separated from RST :
- Yes): Note:4-t
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Environmental Qualification
. o ARI. equipments are qualified to conditions,during an
- ATWSl event up to the' time the ARI: function:is completed)
Yes-8.- Quality Assurance-a) Comply with Generic Lett'er 85-06'
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i Yes,' Note 6" i
- 9... Safety-related power supply a) ARI system power independent;from RTS Yes, Note 5-n b) ARI system can perform'it's function during any j
i loss-of-offsite. power event.
Yes, Note'5 1
- 10. ' Testability at Power a) ARI testable at power Yes y
b) Bypass features conform to bypass criteria used'in RTS Yes j
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f 11.
Inadvertent Actuation
.1 a) ARI Actuation setpoints will not challenge scram-Yes I
b) Coincident logic is utilized in ARI design Yes 12.
Manual Initiation a) Manual initiation capability is provided Yes-
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.' 13. Information Readout a)'Information readout is provided-in main control room
'Yes' u
H 14.
Completion of protective action.once it is initiated-Yes
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. NOTE 1:
The ARI design for LaSalle County Station is safety-related.
It is an i
electrically independent system which is fed from Class IE 125 VDC power i
supplies. The circuit breakers feeding the ARI system have been procured Class 1E.
Relay contacts are used as isolation devices between the safety-related ARI system and the annunciator and/or plant digital computer.
NOTE 2:
The ARI design for LSCS is safety-related with two electrical divisions.
One j
division is ESF Division 1 and the other is ESF Division 2.
is the RPS system which is a four channel electrical arrangement which has individual channel separation. All four RPS channels are routed in separate conduits and never routed with ESF Division 1 or 2 (ARI). The separation
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criteria between RPS subchannels and ESF divisions is further described in 1
l NOTE 3:
I ARI circuits are designed as safety-related circuits.
I NOTE 4:
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l The ARI system is electrically and physically separated from the RTS (RPS).
l See Note 2.
j NOTE 5:
The ARI system is powered from safety-related 125 VDC busses.
The RTS is powered from RPS MG sets (120 VAC).
DC power is available after loss of offsite power events.
NOTE 6:
i The ARI design and installation - treated as safety related and the I
requirements of 10CFR 50 Appendix B are applicable.
i 7914L
-.c CARGENT Q LUNDY,
EN GIN E ERS 1
CHICAGO Commonwealth Edison Company.
-Attachment 2-1 LaSalle County Station - Units 1 & 2 j
1 EVALUATION OF ATWS RECIRCULATION PUMP TRIP (RPT) SYSTEM j
Paragraph (C)(5) of the'ATWS Rule 10CFR50.62 requires a RPT. System as follows:
"Each boiling water reactor'must have equipment to trip the' reactor. coolant-j recirculating pumps automatically under-conditions indicative of a'n ATWS..
This equipment must be designed to perform its function in a reliable manner."'
Each reactor recirculation (RR) pump'at LaSalle County. Station is fed from one.
lL of two sources of power dependent on the operating condition of the' unit.
For.
L low power operation, each pump'is fed from a low frequency motor generator.
(LFMG) set. The input to.the LFMGtis via a-4160 volt Breaker 1A/B and the 1
output is Breaker'2A/B. For full power operation, each' pump is fed' from a:
6900 volt Breaker 3A/B in' series with Breaker.'4A/B..
The recirculation pump speed control logic'is designed to provide either full speed or reduced speed, depending on the conditions prevailing in the reactor vessel.and in the supporting control systems.
The recirculation pump motors are operated at normal or low frequency providing'the full (100%) or. low (approximately 25%) pump speed respectively in response-to the pump control relay logic.
i There are two types of RPT systems at LaSalle County Station.
One is the end of cycle (E0C) and the other, anticipated transient without scram (ATWS).
.i E0C-RPT is the recirculation control system.that trips. the_ recirculation pumps from the 60HZ frequency power supply (6.9KV) in response'to a turbine / generator trip or load rejection.
The function'of'the E0C-RPT System is to reduce the severity of the fuel thermal: transient due to turbine / generator trip or load rejection event-by tripping the recirculation pumps early in the event. The rapid core flow reduction increases void content and thereby reduces reactivity in conjunction with. the control rod -
l-ATWS-RPT is the system that trips the recirculation pumps on low-low reactor water level or high dome pressure to limit the pressure rise in the vessel and reduce reactor thermal power given a failure of RPS to scram. The ATWS: trip-separates the recirculation pump motor from-either of its power supplies.. The
.j initiating signals.are sensed by dedicated high vessel' pressure or low-low water level instruments.
The function of this-action is to mitigate the effects of an unexpected plant transient such 'as turbine trip with ~a failure to scram due to a common mode failure in the reactor protection. system-(RPS)
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trip.
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=SARGENT G LUNDY' 7
.E N G 1 N E E R Sc CHICAGO '
' Commonweal th. Edi son Company LaSalle-County Station.- Units 1 & 2 The ATWS-RPT for each pump consists of Level Instruments.B21-N036A-and.C (Pump.A),=B21-N036B and D (Pump B); Pressure Instruments B21-N045A.and C (Pump A), B21-N045B and D (Pump,B). 'The ATWS RPT logic'for each pump;is:-
RR Pump A: On'e outLof two Level;A or Pressure-A or one out of two Leve1~C or Pressure C RR Pump'B:.One out:of two Level B.or Pressure B
.or one out of two Level:D or. Pressure D' ATWS RPT. logic trips the-single coil of the LFMG set output. breaker if running -
at low power or.the second trip'. coil ofa 6.9KV Breaker ~ 3A(B).if running at full pump speed.- See the attached figure which shows a single 'line representation of the RPT systemscused at LaSalle.
The first' trip coil of-each.6.9KV' breaker is. physically separated from the.second to provide electrical separation for the RPS inputs of the E0C.RPT.. See. Table ~ AL for a '
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summary of the LaSalle BWR 5 ATWS-RPT design.
- The LaSalle County. Station BWR 5 RPT design for ATWS is zvery similar ;to the-original BWR 4. design as described-in-the General Electric. Company Document NEDE 31096-P, Section 4.
This similarity. exists primarily with the' design ~
concepts listed in Table A.
The LaSalle BWR 5 hardware isi significantly different than the original BWR 4 design because of.the useJof the LFMG' set and hydraulic flow control system..The LaSalle design is_'not the same as the Monticello or Modified' Hatch Design described.in Licensing' Topical Report =
Licensing Both the EOC-RPT and ATWS-RPT are described in Apendix'G of the LSCS FSAR.
Specifically, Amendment 31 dated April,1978, Figure G~ A-2, provides -the functional control diagram for the ATWS-RPT, which ~is consistent with the-L logic description above.
FSAR Question 031.208,, Item 3,^has also addressed Figure G.A-2.
The Safety Evaluation Report (SER) NUREG-0519 for LSCS dated March,1981, Section 15.2.1, states, "Early operator. actionLas described above, in conjunction with the recirculation. pump trip whirh has already.been approved and installed at LaSalle, would provide significant protection for-some ATWS events..."
j Generic Letter 85-06 Quality Assurance Guidance for' ATWS Eeuipment The LaSalle County Station ATWS-RPT was designed as functionally'non-safety-related. Major components for the system such'as the associated:switchgear, i
level and pressure instrumentation, 'and cabling were: procuret as Class a 1E/ Safety-related and, subsequently, installed in a quality t/pe manner.
The-major components associated with the LSCS AT.WS-RPT are designated in the Q-List as safety-related. Maintenance of this equipment will.be in accordance with this designation and the-applicable CECO Quality Assurance Controls.
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SARGENT & LUNDY ENGINEERS CHICAGO Commonwealth Edison Company LaSalle County Station - Units 1 & 2 ATWS Rule Compliance The LaSalle County Station ATWS-RPT design, currently. installed, will' perform its function reliably, thereby fullfilling the requirements' of the ATWS Rule l
This conclusion is based on the review of the:.1..
Reactor water level and drywell pressure trip logic,
- 2. Channelization of' logic for each pump,
- 3. Independent trip coils and separation from the E0C-RPT
- 4. Reliable trip coil de power supplies, and
- 5. Independence of the sensors from any RPS circuitry.
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EARGENT.O LUNDY ENOINEERO CHICAGO '
TABLE A 1
LaSalle' County Station BWR 5 ATWS RPT Design' Topic LSCS'BWR 5 1
l Trip Logic 1/2 level or~1/2. pressure l
' to trip each.LFMG ' set output breaker. or'
~
6'.9KV Breaker 3A(B)
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Pump Trip Logic'A & C for "A"
Pump
..i Logic B'& D for "B"
Pump I
Breaker Used for-Trip
- 1) LFMG Output Breaker 2A(B) - 1.' trip-coil:
J 2) 6.9KV Breaker 3A(B) - 2 trip _ coils j
(l' coil used for E0C)'
i Tripping Power Source Breaker 3A(B) or 2A(B) de: control power.
source. - This power is available during loss of off-site power.
Independence from RPS Dedicated sensors used independent from RPS.
B21-N036A, B, C,.& D, level.
2 B21-N045A, B, C, & D pressure.
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- ATTACHMENT 3' IMPLEMENTATION OF 10CFR 50.62(c)(6)-
. REQUIREMENT FOR THE STANDBY LIOUID CONTROL'(SBLC)-
SYSTEM AT5LASALLE' COUNTY > STATION UNITS 11&'2 a
.Commonwea1th.EdksonLCompany(CECO)is1intheproce5s:of: implementing Laodifications for the SBLC System.at-LaSalle Station. Units l and 2 as required -
.by'10 CFR 50.62(c)(4)'.1 This ATWSfrule, as:it, applies lto the.SBLC System,.
requires increased (injection to 86 gpm;of'13 weight' percent,' sodium pentabo
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a rate. solution. CECO.has chosen the two pump SBLC system operation alternative, and followed the conditions set forth in.the Nuclear' Regulatory Commission's-Safety. Evaluation'of Topical Report ~(NEDE-1096-p), " Anticipated: Transient
'Without Scram; Response to NRC.ATWS' Rule,'10,CFR 50'62" to'atisfy the.861gpm:,
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equivalency requirement.:.The SBLC. System at LaSalle County Station ~(LSCS),1 after the modifications have been implemented,'will satisfy:the:" equivalent.
control capacity" requirements outlined in Article 4.0lof'the staff's SBR.
I CECO will perform a modification toithe suction and discharge piping of the SBLC System pumps to implement the ATWS-50.62 rule'after/ performing-H hydraulic calculations. The common suction:line.will be. replaced with a~
l larger pipe and rerouted to preclude NPSH/ flashing-problems ~.Thefdischarge piping will be modified to include. pulsation dampers to attenuateLthe positive displacement pumps' pressure variations andLpiping vibration.. !Thefrelief; valve setpoint will~be raised to-1500lpsig from;1400 psig.due.to the
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increased ' pressure developed by the frictional resistance.::The' Technical Specification'will be revised 1for the relief. valve setpointichange and.the discussion in the bases will be amended to reflect the ATWS modification. The aforementioned modifications will not invalidate the original design basis off the SBLC System. Note that redundant SBLC pumps are not available; (or required) for the ATWS event.
A one-time two pump SLBC System operational test will-be' performed as i
part of the modification testing program. This test will use demineralized.
water to ensure that the appropriate flow rate into the reactor vessel is:
achieved. Surveillance and periodic testing.of the'SBLC System will be n performed in accordance with the original,LSCS technical specification requirements after the one-time two-pump operation test is performed.
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. Pace 2 of Attachment 3 l
1 Generic Letter 85-06 Ouality Assurance' Requirement I
The Standby Liquid contro1' System as presentlycinstalled is a special.
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system functioning as a backup reactivity control: system to shutdown the-creactor. Itlis not required to mitigate;the consequences of'any/ design bases >'
accidents described in the FSARiand;.therefore, does not; fully meet,.nor is.-it-required to meet,'all the design' criteria'of'a safety related system such as;
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-single failure or.electrica1' separation'. The system, however, was; installed-with all the'necessary safety;related Q.A. controls applied-(10CFR50 Appendix.
~l B Q.A. Program)..
u The SBLC.two pump ATWS modification will be installed and maintained, in accordance with'the Commonwealth Edison. Quality Assurance Program as.
described in CECO Topical Report Revision 46, dated 8/11/87.- New components l
purchased for the modification will'be' purchased as safety-related and Code.
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Replacement components will be purchased in accordance.with the original' l
I system procurement. The system will be classified for. Quality Assurance
-i purposes, as, non safety related with safety related~ components..
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-Page 3 of Attachment 3-U r
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JRequest For Additional Information-Provide the following information regarding the proposed two' pump-standby liquid control system--(SLCS)' equivalency requirement:
. Describe how the equivalent' flow requirement will-be' met 1.e., by-increasing the concentration'of-the. solution or the: enrichment of B10,.
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What total two pump flow (gpm) will be achieved?;
Response
The equivalent flow requirement willlbe met by.slightly increasing-the tank concentration from 13.4 percent by weight to 13.6 byLweight. This will compeesate for a slightly lower flow than'86'gpm (82.4 gpm).as called for' in the rule.
The total.two pump flow will be approximately 84 gpm but.in'no cased lower than 82.4 gpm. 82.4 gpm @ 13.6 percent by weight.is equivalent. control:
capacity to 86 gpm @ 13.0 percent by weight.
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