ML20206E862
| ML20206E862 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 04/06/1987 |
| From: | Allen C COMMONWEALTH EDISON CO. |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 2936K, GL-85-06, GL-85-6, NUDOCS 8704130764 | |
| Download: ML20206E862 (11) | |
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CORMINNM9elth Edloon One First Nabonal Plaza. Chca00, Illinois 7 Address Reply to: Post Omco Box 767 b/ Chcago, Illinois 60690 0767 o
. I.
April 6, 1987
.c U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C.
20555
Subject:
LaSalle County Station Units 1 and 2 Implementation of 10 CPR 50.62, ATWS Rule NRC Docket Nos. 50-373 and 50-374 Reference (a): Letter dated December 19, 1987 from A. Bournia to D.L. Farrar.
l Dear Sir-Attached in response to the above noted NRC request is the Commonwealth Edison Company (CECO) response for the LSCS ATWS design compared to the guidelines set forth in the NRC Safets* Evaluation of Topical Report (NEDE-31096-P). Station Nuclear Engineering Department (SNED) has reviewed the attached response prepared by Sargent & Lundy (S&L) and concurs with their evaluation.
5' of the response describes the proposed ARI system design for LaSalle County Station and has indicated that it satisfies the requirements f,
of 10 CFR 50.62 (c)(3). Attachment 2 describes the proposed SBLC system design and its conformance to the requirements of 10 CFR 50.62 (c)(6).
l The Reactor Coolant Recirculating Pump Trip (RPT) was not included in this response. The RPT (ATWS) fix was installed during the pre licensing (construction phase) period at LaSalle County Station.
In order to review the existing RPT design and determine its conformance with the requirements of the ATWS rule, CECO requires an additional three weeks r4 piavide this evaluation.
In response to the Quality Assurance CJi Av ir the ATWS Equipment J>
(Generic Letter 85-06) the following would be av Aica~ i;
,i 1.
ARI design and installation - treated as safety related and the requirements of 10CFR 50. Appendix B would be applicable, i
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4 US NRC April 6, 1987 2.
SBLC and RPT design and installation, will be addressed in the submittal on RPT prior to the end of April.
If you have any additional questions regarding this matter, please contact this office.
Very truly yours, I
l 1
V C. M. Allen Nuclear Licensing Administrator 1m Attachments cc:
Dr. A. Bournia - NRR Regional Administrator - RIII NRC Resident Inspector - LSCS M. C. Parker - IDNS 2936K
4 ATTACHMENT 1 IMPLEMENTATION OP 10 CPR 50.62 REQUIREMENTS FOR THE ALTERNATE ROD INJECTION (ARI) SYSTEM AT LASALLE COUNTY STATION - UNITS 1 AND 2 REOUIREMENT - 10 CPR 50.62(c)(3):
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 at LaSalle County Station (LSCS) - Units 1 and 2 satisfies the requirements of 10 CPR 50.62(c)(3). The following text correlates with the NRC SER Appendix A checklist enclosed with Attachment 1.
1.
ARI System Punction Time The ARI system function will begin within 15 seconds and be completed-within 25 seconds. The results of that ARI transient analysis described in reference 4 indicate that the ARI function time objectives are met as follows:
The ARI is initiated 15 seconds after a recirculation pump trip on high vessel pressure (1135 psig) and is completed within 3.67 seconds. The total time from MSIV closure start to the completion of the ARI scram is 21.88 seconds.
b) Turbine Trip (TTWB)
The ARI is initiated 15 seconds after a recirculation pump trip on high vessel pressure (1135 psig) is completed within 3.66 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 where the suppression pool temperature reaches 110*P.
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 ARI scram completion is 18.90 seconds.
a e 2.
Safety-Related Requirements The ATWS Rule does not require the ARI system to be safety grade; however, the implementation of ARI must be such that the existing protection system continues to meet all applicable safety-related criteria. Specifically, qualified isolators should be used for ARI system interfaces with safety systems.
The ARI design at LSCS - Units 1 and 2 is safety-related with two electrical divisions. One division is ESF Division 1 and the other is ESP Division 2.
The equipment at the interfaces is qualified to IEEE Standards 323 and 344 and divisional physical separation and independence is maintained, no isolation devices are required at the ARI/ Safety-System interfaces.
Interfaces between the ARI and non-safety related readouts are provided by Class 1E relays where the safety-related relay coil is divisionally 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 header exhaust valves and perform a function redundant to the backup scram system. ARI self-redundancy is not required.
The LaSalle design employs redundant dual-coil solenoid-operated valves to 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 and drain valves permitting those valves to close. The one-out-of-two taken twice logic of either ESP 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 should 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 to 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, I
b) DC powered Valves The solenoid valves used are powered from the respective divisionally associated 125 Vdc ESP buses (1 or 2) 1 l
e
.. c) Instrument Channel Components The ARI channel components employ linear analog transmitters and trip units. The level 3 and reactor pressure vessel high pressure reactor trip function of the reactor protection system employ differential pressure and pressure switches in their channel logic. Therefore, the ARI conforms to the ATWS Rule equipment diversity requirements.
5.
Electric Independence from the Existing RTS a) The ARI design for LSCS is safety-related with two electrical divisions; one division is ESF division 1 and the other being ESP division 2.
The RTS at LSCS is the Reactor Protection System (RPS),
which is a 4-channel electrical arrangement which has individual channel separation. All four RPS channels are routed in separate conduits and never routed with ESF division 1 and 2 (ARI). The separation criteria between RPS subchannels and ESP divisions is further described in Chapter 8 of the LSCS FSAR.
b) The ARI system for 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 di i.gned to be physically separated from the RTS (which is the RPS system at LSCS). The safety-related design of this system with ESP separation criteria ensures that physical separation is maintained.
7.
N/A 8.
N/A 9.
Safety-Related Power Supply 1
a)' The power for the RTS (RPS) is 120 Vac from the RPS motor generator sets. The power for ARI is from 125 Vdc safety-related buses, b) The 125 Vdc buses remain energized during loss of offsite power events.
10.
Testability at Power The ATWS Rule guidance stated that the ARI system should be testable at power. The LaSalle design conforms to the Rule as follows:
. 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) level 2 "or (inclusive)" one ot 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 ESP 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 without disabling the ARI system.
b) Indication of Bypass Test switches and (white) indicating lights are provided for testing the ARI control logic. These lights provide test status indication in the control room at panel H13-P603. A red indicating light is 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.
Guide 1.47 is identical to that which was accepted by the NRC for the safety-related reactor trip system, i.e.,
in accordance with the applicant /AB 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 Reg.
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 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 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.
e,
b.
Coincidence Logic The coincidence actuation logic described previously in 10(a) is also intended to prevent 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, i.
' 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 specific ARI system features identified in the l
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.
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APPENDIX A CHECKLIST FOR PLANT SPECIFIC REVIEW OF ALTERNATE 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 b) Class lE isolators are powered from a Class IE source Yes, Note 1 c) Isolator qualification documents 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 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 L
physical separation from the existing RTS (a) ARI system is physically separated from RST Yes, Mote 4 7.
Environmental Qualification ARI equipments are qualified to conditions during an ATWS event up to the time the ARI function is completed Yes 8.
Quality Assurance a) Comply with Generic Letter 85-06 Yes 9.
Safety-related power supply a) ARI system power independent from RTS Yes, Note 5 b) ARI system can perform its function during any loss-of-offsite power event.
Yes, Note 5 10.
Testability at Power a) ARI testable at power Yes b) Bypass features conform to bypass criteria used in RTS Yes 11.
Inadvertent Actuation a) ARI Actuation setpoints will not challenge scram Yes b) Coincident logic is utilized in ARI design Yes 12.
Manual Initiation a) Manual initiation capability is provided Yes 13.
Information Readout a) Information readout is provided in main control room Yes 14.
Completion of protective action once it is initiated Yes t
. NOTE 1:
The ARI design for LaSalle County Station is safety-related.
It is an electrically independent system which is fed from Class IE 125 VDC power supplies. The circuit breakers feeding the ARI system have been procured Class 18.
Relays 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 division is ESP Division 1 and the other is ESF Division 2.
The RTS at LSCS 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 rated with ESF Division 1 or 2 (ARI). The separation criteria between RPS subchannels and ESP divisions is further described in Chapter 8 of the LSCS FSAR.
NOTE 3:
ARI circuits are designed as safety-related circuits.
NOTE 4:
The ARI system is electrically and physically separated from the RTS (RPS).
See Note 2.
i 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.
1 2936K f
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l ATTACIWWIT 2 IMPLEMENTATION OF 10 CPR 50.62(c)(G) i REQUIREMENTS FOR THE STANDBY LIOUID CONTROL (SBLC) SYSTEM AT LASALLE STATION UNITS 1 AND 2 i
Commonwealth Edison Company (Ceco) is in the proccca of implementing modifications for the SBLC System at LaSalle Station Units 1 and 2 as required by 10 CFR 50.62(c)(4). This ATWS rule, as it applies to the SBLC System, requires increased injection to 86 gpm of 13 weight percent, sodium pentabo-rate solution. CBCo 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-31096-P), " Anticipated Transient Without Scram; Response to NRC ATWS Rule, 10 CFR 50.62" to satisfy the 86 gpm equivalency requirement. The SBLC System at LaSalle County Station (LSCS),
after the modifications have been implemented, will satisfy the " equivalent j
control capacity" requirements outlined in Article 4.0 of the staff's SER.
Ceco will perform a repair to the suction and discharge piping of the SBLC System pumps to implement the ATWS-2A rule after performing hydraulic calculations. The common suction line will be replaced with a larger pipe and rerouted to preclude NPSH/ flashing problems. The discharge piping will be modified to include pulsation dampers to attenuate the power pumps' pressure variations and piping vibration. The relief valve setpoint will be raised to 1500 psig from 1400 psig due to the increased developed pressure caused by the frictional resistance. The Technical Specification will be revised for the relief valve setpoint change and the changes to the bases of the Technical 4
Specification. The aforementioned modifications will not invalidate the original design basis of the SBLC System. Note that redundant SBLC pumps are i
not available for the ATWS event.
A one-time two-pump SLBC System operational test will be performed using demineralized water to ensure that the appropriate flow rate into the reactor vessel is achieved. Surveillance and periodic testing of the SBLC system will revert back to the original LSCS technical specification requirements after the one 'ime two-pump operation test is performed.
2936K 4
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