ML20126J078

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Forwards Revised Response to GL 91-11, Resolution of GI 48, LCOs for Class 1E Vital Instrument Buses & 49, Interlocks & LCOs for Class 1E Tie Breakers Per 10CFR50.54(f)
ML20126J078
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 12/30/1992
From: Beckman W
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
REF-GTECI-048, REF-GTECI-049, REF-GTECI-NI, TASK-048, TASK-049, TASK-48, TASK-49, TASK-OR GL-91-11, NUDOCS 9301060027
Download: ML20126J078 (28)
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William t. Beckrnan ihnt Manager A8KNNiAN'E MROGRE55; Big Rock Point Nuclear Plant,10269 US 31 North, Charlevoix, MI 49720 December 30, 1992 Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 DOCKET 50-155 - LICENSE DPR BIG ROCK POINT PLANT -

REVISED RESPONSE TO GENERIC LETTER 91-11:

RESOLUTION OF GENERIC ISSUES 48, "LCOs.

FOR CLASS lE VITAL INSTRUMENT BUSES," AND 49 " INTERLOCKS AND LCOs FOR CLASS-lE TIE BREAKERS" PURSUANT.T0 10 CFR 50.54(f)

Pursuant to Section 50.54(f) of Title 10 of the Code of Federal Regulations and Section:182 of the Atomic Energy Act, the_ original submittal dated January 29,.

1992, provided the certification that appropriate procedures and administrative.

controls conforming to the guidance provided in Enclosure 1 to the Generic Letter

'had been implemented.. However, after review, the Office of Nuclear Reactor Regulation (NRR) expressed a concern through the NRR Project Manager that there were no. procedural time limitations (operability requirements) placed on the

. inverter or other onsite power sources that supnly 120 V At to the Reactor Protection System Bus No. 3.- Per a conference call between NRR and Big Rock Point Staff ~on December 1,-1992, a revised response was agreed upon to provide F

~the. basis for this position.

} Ybw' 1 am L Becknian Plant Manager CC: ; Administrator, Region III, USNRC NRC Resident Inspector Big Rock Point

' ATTACHMENT 040063 (1

9301060027 921230'

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u CONSUMERS POWER COMPANY-BIG ROCK POINT PLANT Docket 50-155 - License DPR-06 At the request of the Commission and pursuant to the Atomic Energy Act of 1954-and the Energy Reorganization Act of 1974, as amended, and the Commission's Rules and Regulations thereunder, Consumers Power Company submits our revised response to NRC letter dated July 17, 1991, entitled, " Response to Generic letter 91-11; Resolution of Generic Issues 48, LCOs for Class lE Vital-Instrument Buses, and 49,-Interlocks and LCOs for Class IE Tie Breakers, Pursuant to 10 CFR 50.54(f)".

Consumers Power response is dated December 30, 1992.

CONSUMERS POWER COMPANY To the best of my knowledge, information and belief, the contents of this submittal are truthful and complete.

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PC-BY David P Hoffman, Vice q t Nuclear Operations Sworn and subscribed to before me this 30th day of December 1992.

(dat4 M v, #

&ver/7 g, Aa,y.

, Not'ary Public

,faas,n County, Michigan My commission expires d"cem6er 3, /##4

( SEAL )

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i ATTACHMENT-1 Consumers Power Company Big Rock Point Plant Docket 50-155 REVISED RESPONSE TO GENERIC LETTER 91-11 4

December 30, 1992 6 Pages

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j ATTACMENT 1"

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-y REVISED RESPONSE TO GENERIC LETTER 91_1F ei H

zi 1.0 SYSTEM DESCRIPTION-Class IE Buses j

i 480 V Emergency Bus MCC-2B-~

Vital Instrument Buses 115 V Reactor Protection System Bus 1 115 V Reactor Protection System Bus 2 120 V Reactor Protection System Bus 3 Vital Instrument Panels Panel lY Panel 2Y-Subpanel 2Y Panel 3Y Inverters Topaz Static Inverter; RPS Bus 3 normal supply Rod Position Indication Motor-Generator set Tie Breakers Connectino Redundant Class IE Buses at Bia Rock Point" None Discussion The Big Rock Point Station Powe'r System has a sinale Class lE Motor Control Center (MCC-28) Bus.

This 480 V Emergency Bus-is normally-powered.by Motor Control Center 2A via the Bus 2A-2B tie through breakerc n

52-2A28. MCC-2A-draws its power from the 480 V; Load Center Bus 2 via:

breaker: 52-2A,- and in turn =is powered by: the 2400 V Station Power Bus;via-1 Station Power Transformer (2400/480V) No. 22.. An alternate. power supply?

-path for the MCC-2B Emergency Bus exists'from the'2400 V Station Power Bus via the Station Power Transformer (2400/480V) No..ll, the 480 V Load L

Center -Bus.1, breaker 52-1A, the Motor Control Center IA,;and along the Bus IA-28 Tie'via breaker:.52-1A28. Note that both Station Power Transformers are. energized.by-the same 2400 V.switchgear whichcis supplied by a voltage regulating transformer through-Station Power?

h Transformer No. l_from the unit generator,Jthe 138-kV' grid orithe 46 kV' ll grid.

In the event. that the previously described paths are unavailable, the MCC-28: Emergency Busicanibe fed f om the' Emergency; Diesel Generator.

~or the. Standby. Diesel _ Generator.-

The ll5.V Reactor Protection System (RPS) Buses 1 and 2 are ?owered by separate motor generator:(MG) sets. The motor inputs are ta(en from7 MCC-1A and MCC-2A respectively.

The MG_ sets are' mechanically coupledito inertia l flywheels,- enabling them to ride out minor system voltage -

I disturbances. A complete loss _ of supply for 10 seconds can be. tolerated,

- by 'the sets without-causing reactor protection. system operation due to-i

l 2

Ml&GBENL1 REVISED RESPONSE TO GEhfJl0 LETTER 91-11 low voltage. The 120 V Reactor Protection System Bus 3 (Neutron Monitoring Bus) is normally powered by an inverter from the 125 V DC Distribution Panel No. 1.

An alternate 115 V supply can be switched to either of the two Reactor Protection System Buses or to the Neutron Monitoring Bus No. 3 via Panel lY. This alternate supply is interlocked so that only one of these three buses can be supplied at any one time by the alternate power supply.

Panels lY, 2Y, subpanel 2Y and 3Y are normally. fed by MCC-1 A.

If MCC-IA is lost, an automatic throwover operates to supply power from the MCC-2B Bus. Refer to Attachment 4 for the loads carried by Panels lY, 2Y, subpanel 2Y and 3Y.

The Rod Position Indication System is normally powered by Panel lY, but can be backed-up by a motor-operated set (inverter) that starts automatically upon loss of power to Panel lY. This motor-generator set (inverter) is fed by the 125 V DC Distribution Panel No. 1.

The vital instrument buses and panels are unique to themselves; there is no redundancy in instrumentation between them. of this submittal is a simplified diagram of the Big Rock Point Station Power System. Attachment 3 represents the Reactor Protection System Buses.

2.0 GENERIC LETTER RECOMMENDED ACTION A.

Time limitations and Surveillance Reauirements for Vital Instrument Buses and Panels Time Limitations At Big Rock Point, the vital instrument buses and panels are indirectly required by Technical Specifications to be operational at all times during power operation and refueling.

Limiting Conditions of Operation are determined by the affected instruments / controls associated with the bus / panel. There are no limitations associated with inoperability of vital instrument buses or panels because of this design.

Surveillance Reauirements Tests are performed in accordance with Technical Specifications to insure the operability of the instruments / controls associated with the vital buses / panels.

B.

Ilme Limitations and Surveillance Reauirements for Inverters and Other On-Site Power Sources to the Vital Instrument Buses and Panels Big Rock Point Technical Specifications and procedures contain time limitations and surveillance requirements concerning the major on-

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3 ATTACHMML1 REVISED RESPONSE TO GENERIC LETTER 91-11 site power sources to the vital instrument buses and panels (refer-to Attachment 5).

flod Position Indication Motor-Generator Set To ensure the operability of the Control Rod Fosition Indication System, a motor-generator set (inverter) is used to backup the normal lY power supply. Operability is tested monthly using Operations Surveillance Procedure T30-05, Monthly Operational Test of Control Rod Position M-G Set.

The power source for the motor-generator set is the 125 V DC station batiery.

Technical Specifications require the Station Batteries to be operable under all conditions except cold shutdown, therefore directly affecting the operability of the inverter.

Topaz Static Invertet To ensure the operability of RPS Bus 3, two static power inverters are available.

One is installed and operable, and the other unit is stored on-site as a spare. The power source for the inverter is the 125 V DC-station battery.

Technical Specifications require the Station Batteries to be operable under all conditions except cold shutdown, therefore directly affecting the operability of the inverter and ultimately RPS Bus 3.

If the inverter is removed from service, the operator can transfer the RPS Bus 3 power source to the alternate 120 V AC power supply Panel lY in accordance with Standard Operating Procedures.

Each refueling outage these inverters are exchanged, and the removed inverter is sent to an offsite lab for testing. The tested inverter is then usually returned to the site before power operation, however there is no requirement to do so.

Recommended Action oer Generic letter 91-11' Ensure that your plant has procedures that include time limitations and surveillance requirements for inverters or other onsite power sources to the vital instrument buses.

If plant procedures do not include time limitations for inverters or other onsite power sources to the vital instrument buses, the basis for such a position must be adequately-evaluated.

The evaluation should address existing regulations and plant design bases, and should specifically demonstrate that adequate consideration has been given to the possibility of loss of offsite power that coincides with a worst-case additional single failure.

In addition, l

the analysis should consider the time delay required for the emergency generators to pick up loads, because in typical plants, if an inverter serving a vital instrument bus is not available, a loss of offsite power -

will cause numerous actuations because of the delay time while the diesel I,-

generators are starting. Therefore the analysis should also consider i

malfunctions that-do not always have a preferred failure mode, (e.g.,

instrumentation or controls that initiate a switch of emergency core F

cooling from injection to recirculation or initiate isolation of the steam generators).

If the alternate power sources for the vital buses l-

=-

4 ATTACHMERL1 REYlSED RESPONSE TO GENERIC LETTER 91-11 cannot receive power from the diesel generators, the evaluation should include this condition.

Evaluation 1.

Existina Reaulations and Plant Desian Bases Time limitations (operability requirements) for the inverter or other alternate power sources that supply the 120 V O RPS Bus 3 are not included in plant procedures, which is contrary to the recommended actions per Generic letter 91-11. Therefore, the basis for such a position is required to be evaluated.

Big Rock Point was evaluated by the Systematic Evalus e i Program (SEP) which was initiated in February 1977 by the US s

. ear Regulatory Commission to review the designs of older operating-nuclear reactor plants to reconfirm and document their safety. As you are well aware, Big Rock Point was licensed prior to the publication of 10 CFR 50, Appendix A, " General Design Criteria" (GDC) for Nuclear Power Plants", specifically GDC 17, 21, 34 and 35.

The SEP review identified topics that related to Electrical Power Reliability, and Big Rock Point responded with-satisfactory -

completion of these topics documented by the NRC Staff in NUREG-0828, Integrated Plant Safety Assessment - Systematic Evaluation Program; Big Rock Point Plant, dated May 1984.

Systems required to bring the Big Rock Point nuclear reactor from hot shutdown to cold shutdown with only offsite or onsite power available with a single failure meet the current NRC criteria.

However, the vital indications in the control room such as reactor pressure, temperature and level indicators can be lost given a single failure. The initial recommendation of the NRC Staff was that a design modification be incorporated such that independent and redundant indication of the process variables that are vital for the safe shutdown of the reactor are available in the control room.

This requirement was analyzed in SEP Topic VII-3, Systems Required For Safe Shutdown, and the following conclusion was reached:

"There are two facts relevant to this issue.

First, in the current Big Rock Point design there is only one dedicated diesel connected to the 480 kV emergency bus (a second diesel generator is available, however credit is not taken for its use within the first three and a half hours following an accident sequence initiator).

Given a total loss of AC power which implies loss of vital instrumentation in the control room, the plant can be shutdown and cooled down for at least four hours.

(This is due to the fact that the emergency condenser has sufficient inventory to cool the reactor for four hours and the only normally closed valves in this system are DC powered and have control room indication. This offers a reasonable period of time to try to recover

5 ATTACHMGL1 REVISED RESPONSE TO GENERIC LETTER 91-11 either offsite power or the onsite emergency power.)

Second, a failure that would affect only the-shutdown panel does not affect the operability of the safety systems required to shutdown Big Rock Point.

Based on these facts, addition of a redundant vital instrumentation panel does not seem to offer much in the way of reduction of risk due to operation of this plant."

Additionally, the issues in Regulatory Guide 1.97 (Detailed Control Room Design Review and the Safety Parameters Display System) as they apply to Big Rock Point have been resolved, and are dc.umented in NRC Safety Evaluation to Consumers Power Company dated 7/11/90 and additional letter 8/2?/90.

2.

Time Delay for EDG to Pick Uo loadj The only instrument fed from RPS Bus 3 is Neutron Monitoring Channel-3.

If RPS Bus 3 becomes inoperable, a "Downscale Alarm" will be annunciated in the control room and a control rod withdrawal-block -

will occur (this function is tested every 30 days using Reactor Protection System Surveillance Test T30-01).

NOTE:

In accordance with Technical Specifications, any one of three power range monitors may be taken out of. service for surveillance testing or maintenance during reactor operation for the reason that a trip on either of the two remaining channels shall scram the reactor. There is no time limitation associated with this condition.

There are no Engineered Safeguards Features (ESF) associated with Channel 3 (other than RPS logic described above) that would be affected after a loss of offsite power coinciding with a worst case additional single failure as described in SEP Topic VII-3, therefore

" malfunctions that do not always have a preferred failure mode" are not applicable in the operation of RPS Bus 3.

Furthermore, even though the alternate power supply, Panel lY, for RPS Bus 3 can receive power from the Emergency Diesel Generator (EDG), it is unlikely that the operator would power RPS Bus 3 from the alternate source, if that source was receiving power from the EDG..The alternate power supply is common to the three vital instrument

.I buses, and is designed to be manually controlled from the control room'. This alternate power supply controller is interlocked so that only one of these three buses can be supplied at any one time from Panel lY. Station procedures direct the-operator to transfer RPS I

Bus 1 to the alternate power supply if a loss of station-power has occurred. This bus offers the' operator Neutron Monitoring Channels 1 (wide range) and 7 (source range), four out of eight incore monitors and an air ejector off-gas monitor. Therefore the time delay associated with a EDG starting and picking up load is inconsequential in regards to RPS Bus 3.

6 ATTACHMERL1 -

REVISED RESPONSE TO GENERIC (ITTER 91-11 Position Summary RPS Bus 3 only provides an additional neutron monitoring channel; there are no ESF actuations associated with this bus.

If RPS Bus 3 is inoperable, a trip on either of the two remaining channels will scram the reactor.

During a loss of offsite power (EDG or Standby Diesel Generator operating) other means exist to determine flux levels in t;,a reactor.

In the control room, reactor pressure, level, control rod -

position indications are available.

In the Alternate Shutdown Building (Appendix R), completely separate from the control room, steam drum pressure and level indication is available-(EDG not required) to indirectly monitor the increase or decrease of heat production in the reactor core.

(Consumers Power Company identified no need for flux / power level indication in the Alternate Shutdown Building in a letter submitted to the NRC dated September 24, 1981.

By a NRC Safety Evaluation dated March 8,-1983, the NRC concurred with Consumers's position.) The Alternate Shutdown batteries-can support connected loads for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> following a loss of offsite.

power if the EDG is unavailable.

For these reasons, including time limitations on the Topaz Static inverter and other onsite power sources to the 120 V AC RPS Bus 3 in plant procedures is not warranted.

C.

Time Limitations and Surveillance Reouirements for Tie Breakers That Connect Redundant Class lE Buses at One Unit or That Can Connect flass lE Buses Between Units at the Same Site Not applicable to Big Rock Point, because redundant buses are not utilized in the licensed design.

3.0 CONCLUSION

The primary objective of Generic Letter 91-11 is to verify that plants.

are not being operated in violation of apolicable regulations. Consumers.

Power Company is confident that'the electrical configurations, administrative controls and surveillance procedures at. Big Rock-Point' conform to existing SEP regulatory requirements.

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ATTACHMENT 2 Consumers Power Company Big Rock Point Plant Docket 50-155 STATION POWER SYSTEM-Revised Response to Generic Letter 91-11 December 30, 1992

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T ATTACHMENT 3 Consumers. Power Company Big Rock Point Plant Docket 50-155-REACTOR PROTECTION SYSTEM BUSES Revised Response to. Generic Letter 91-11 December'30,;1992 1 Page l~

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ATTACHMENT 4 Consumers Power Company Big Rock Point Plant Docket 50-155 VITAL PANEL LOADS Revised Response to Generic Letter 91-11 December 30, 1992-5 Pages i

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l TABLE 11 120/208 VOLT PAWEL 1Y (TURBINE BUILDINC - Alt COMPRESSOR / ELECTRICAL EQUIPMENT ROOM 105)

(SCHEME 1503)

.E-009 21 SWITCH NO.

DESCRIPTION (LEFT-HAND SIDE)_

SCHEHR_ DRAU!NC WO. -SHEET-(Haln Lug) 1Y Teed from HCC-1A, Position 52-1 A53 1501 E-101 1

i Backup feed From HCC-2Be 1502 E-101 1

Position 52-2823 1

2400Y Sultchgear Relay Test 1505 3

Hydrogen Control Panel 5

reedwater System Valve controls 2501 E-109 1

7 Condensate System Valve Controls 2502 E-110 1

9 Diesel Fire Pump 3501 E-114 1

11 Spare 13 Turbine Bypass Valve controls 9502 E-107 1

15 Reactor Yessel Instrumentatlon ID10 C30739-2 6001 C30739 2-6803 C30739 1

6807 C30739-2 17 Area Honitoring 6952 F30762 1

Turbine and Sphere Can Remote Alarm 6956 E-115 1

High Radiation Vent Isolation Logic 8511 E-114 2

19 Reactor Control Rod Positioning 6511 F30731 1

6512 F30731 2

21 Neutron Chamber Positioning System 6516 C30739 2

6517 C30739

.2 23 Reactor Building Level Indication 6503 E-114 1

i 25 Air Ejector Valve Controls 7502 E-115 1

27 Air Ejector Off Cas Monitor Valves and Timer 29 Condensate Demineraliser Control Panel 4504 02 Analyaer -

4505.

E-110 :

1 31 Control Panel C26 1512-E-11 33 Turbine Trip and Test controls

' 6521 C30739 9503 E-107 35 Manual Loader Bypass Viv Rod Hydraulle System 37 Area Monitor Coollag Power Supplies l

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a 120/208 VOLT PANEt,1Y (TURB1HE Bull.DIWC - AIR COMPRESSOR /ELECTRICA1, EQU1PHEWT R00H 105)

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E-009 21 SWITCH HO.

DESCRIPTION (RICHT-HAWD SIDE)

SCH EH E ___ DRAWING HO.

SHEET 2

Control Roors Relay Test 1507 C30708 4

Turbine Selsyns 9501 E-107 1

6 Exciter Switchgest 1509 C30141 8

Control Panel 002 Process Recorders F12325 E-104 1

FI2326 E-104 1

FI2327 E-104 1

FI2328 E-104 1

9801 E-107 1

9802 E-107 1

9901 E-107 1

9902 E-107 1

2801 E-!.09 1

2802 E-110-1 2901 E-110 1

E-110 1

E-105 1

E-106 1

6702 C30734 3

6804 C30739 1

10 Reactor Building Transmitters 8505 E-117 1

8512 E-114 2

8507 E-117 1

12 Shutdown System Valve Control Auxiliary 6504 E-112 1

14 Transformer Deluge 1.ocal Bell Alarms 1702 H740-C19 2

16 Process Radiation instruments 6951 18 Radiation Recorders 8507 E-117 1

6805 C30?t9 1

6954 C30)J9 1

C30739 2

6809 C30739 2

6810 C30739 2

6934 F30760 1

6935 F30760 1

6936 F30760 1

6937 F30760 2

6938 F30760 2

6941 F30761 1

6942 F30761 1

6943 r30761 1

6944 F30761 1

6945 F30761 1

6946 F30761 1

6947 F30761 1

6948 F30761 1

6952 F30761 2

C32012 1

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DESCRIPTION (RICHT-HAND SIDE)

SCHEME _ -DRAWINC WO.

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'20 Reactor Protection System Standby Power 6514 C30743' 1-22 Radwaste control Panel C06-7503 E-11$'

l-7804 E-115 1

7802 E-115

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24 Radvaste System Valve Controls 7501-E-115 1-26 RDS System Panel C40 4501 E30878 <

28 Hakeup Demineralizer Control' Panel C27-30 Security Bullding Receptacles 32/34 Michigan Bell Company Equipment 36 Plant Radio Equipment-38 Domestic Water Accumulator Controller

- 5409 E-113 J

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_120/208 VOLT _PAWEL 2Y (REACTot BUILDlWC - PERSONNEL I4CK' AREA R00N 447)

(SCHEdt 1504)

E-009 22

$ WITCH NO DESCRIPTION (LEFT-if AND SIDE)

SCHEME _ DRAWIWC WO.

SHEET-2 (Main Lug) 2Y feed from HCC-1 A, Position 52-1 A53 ~

1501 E-101 1

Backup Feed from HCC-28, 1502 E-101 1

Position 52-2823 1

Reactor Coo 1Ing Water Auxillary 5501 t-113 1

6509 t-115 2-6811.

3 Control Panels C20 and C22 C30739-i 6802 t-113 1

6812 E-111 1_

8504 E-117 1.

8506-t-117 1

5 Cleanup Demineralizer control Panel 1510 t-ilt 6801-E-);2 ~

1 7

Cleanup Deminerallrer Control valves 1

6505 F 412 1

6507 4-115 2

9 Polson Syst, Shutdown Syst, CRD Pump Press Ind -

6506 t-112 1-11 Control Rod Temperature Indication 6522 C30739 3

13 Control Rod Transmitters 68f C30739 2

15 Recirculation Pump Room Deluge Valve 37M

.E-114 2

SWITCH ?!O.

DESCRIPTION (RICHT-HAND SIDE)

SCHEME _DRAV!WC Wo. SHEET 2

Reactor Building Crane interlocks 6520 E-111

-1 4

Reactor Building Continuous Air Monitor (CAN) 6 Reactor Shutdown System Interlock Valva 6501 t-112 1

8 Emergency Condenser Hakeup Valve 6502 E-112 1

10 Emergency condenser Level Transa LT-3150 6510 2-112 1

12 Rectreulating Pamps and teactor Transmitters 6806 C30739 1

14/16 120/208 Volt Feed to 2Y Subpanet TABLE 14 124/208 VOLT 2Y SUBPANEL (REACTOR BUILDING - PERSONNEL LOCK AREA ROOH 447)

E-009 22 SWITCH'HO._

DESCRIPTION (LEFT-HAND SIDE)

SCHEME DRAWIWC Wo.

SHEET (Hain Lug) 2Y Subpanet Feed From Panel 2Y, Position 14/16~

17 Spare 19-Spare 21 Spare 23 Spare SWITCH NO.

DESCRIPTION (RICHT-HAND SIDE)

SCHEHE

_DRAVINC No. - SHEET

, 16 Area Radiation Honitor (ARH) RE-8258 6952 F30762 1

18 Spare 20 Spare 22 Spare

- - ~ -

~

Pa6] 5 of 3 kMACHMENT IL TABLE 12 120/208 VOLT PANEL 3Y (TURBINE BUILDING - AIR CCHPRESSOR/ ELECTRICAL EQUIPHENT ROOH 105)

E-009 29

_CWITCH NO.

DESCRIPTION (LEFT-HAND SIDE)

SCHEHE DRAWIKO NO.

SHEET (Hain Lug) 3Y Feed From HCC-1 A, Position 52-1 A53 1501 E-101 1

Backup Feed from HCC-28, 1502 E-101 1

Posittoa 52-2B23 1

Reactor Bldg Wator Level Transmitter LT-3171 6821 E-104 1

Reactor Bldg Pr. essure Transmitter PT-200 6815 E-104 1

3 Reactor Core Spray Flow Transmitter FT-2162 FI2327 E-104 1

Reactor Enc'iosure Spray Flow Tranam FT-2164 FI2325 E-104 1

5 Reactor Core Spray Flow Backup FT-2163 F12328 E-104 1

Reactor Enclosure Spray Flow Backup FT-2161 FI2326 c-104 1

7 Relief Valve Honitoring System t,813 032001 1

9 High Range Healtors RI-8322 and R!-8323 11 Reactor Bldg Water Level Transmitter LT-3175 6817 E-104 1

Reactor Bids Pressure Transmittar PT-201 6816 E-104 1

SWITCH NO.

DESCRIPTION (RICHT-HAND SIDE)

SCHEHE DRAWINC NO._

SHEET 2

Core Spray Line Pressure PT-186/PI-412 5801 E-104 1

Reactor Pressure PT-IA07C 6803 C30739 1

4 Pipe Tunnel Damper Position Indication 8510 C31040 1

6 Access Control System (Card Readers) 8 Spent Fuel Pool Level Indication 5803 E-104 1

10/12 Fire Detecflon circuit 3702

.E-114 3

i

7-7.

b

-4 ATTACHMENT 5 Consumers Power Company Big Rock Point Plant Docket 50-155 APPLICABLE TECHNICAL SPECIFICATIONS SECTIONS Revised Response to Generic Letter.91 Decemtie 30, 1992-8 Pages I

i w

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s-v e

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1-r Y'w -

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  • Yr" "

t Limiting Conditions for Operation Surveillance Requirements 11.3.5.3 DERGENCY PO6TR SOURCES 11.4.5.3 DERCENCY PObTR SOURCES

(

Applicability:

Applicability:

Applies to the operational status of the Applies to the periodic testing require-emergency power sources.

ments for the emergency power sources.

Obj ective r Objective.

To assure the capability of the~ emergency To assure the operability"of the emergency power sources to provide power required pcwer sources to provide emergency power for emergency equipment in the event of in the event of a Loss of Coolant Accident.

a Loss of Coolant Accident.

3p g f

}

Specification:

A.

The emergency power system surveillance h

A.

For all reactor operating conditions vill be performed as indicated below.

e except cold shutdown, there shall nor-In addition, components on which reinte-d mally be available one 138 kV line, nance has been performed will be tested.-

F one 46 kV line, one diesel generator systen, one station battery system, 1.

During each operating cycle -

four RDS uninterruptible power supplies

/

including batteries and one siternete (a) Test of automatic initiation

/

shutdown battery system, except sensors and Ioed test the

)

se specified below.

einergency diesel to 180-200 kV generator output 1.

Refueling operations and related for at least 20 minutes.

tearing may be conducted with the 138 kV line de-energized.

(b) Test and calibrate the following instruments and 2.

The 46 kV line or the diesel gen-controls associated with erator may be out of service for diesel generators repair for periods up to three (3) days during reactor operation (1) Fuel oil level, and for extended periods during (2) Oil Pressure tripping.

refueling or shutdown operations.

(3) Water temperature tripping.

'f7 146 Amendment No. If. 97,1 May 31. 1989 TSB0689-0097-NLO4 e

r

_7_,

l A t-Limiting Conditions for Operation Surveillance Requirements 18.3.5.3 EMERCENCY POWER SOURCES (Contd) 11.4.5.3 EMERGENCY POWER SOURCES (Contd) 3.

The diesel generator fuel supply (4) Overspeed tripping.

shall be adequate for three-day

- operation.

(5) Battery undervoltage alarm.

4.

If Specifications A.2 or A.3 are (c) Verify the automatic transfer not met. a normal orderly shutdown of station power free the shall be' initiated.within one (1)

~138 kV line to the 46 kV line.

hour and the reactor shall be shut down as described in Section 1.2.5(a)

(d) Verify the automatic transfer of within twelve (12) hours and shut power sources for the 1Y and 2Y down as described in Section 1.2.5(a) instrument and control panels.

and (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During refueling (e) Verify the cells cell plates.

operatione cease.all changes and battery racks show no visual which could affect reactivity.

indication of physical damage or abnormal deterioration for i

5.

The station battery system and the station battery the RDS p

/

alternate shutdown battery system batteries and the alternate

/

shall be operable under all conditions shutdown battery.

/-

except during cold shutdown. If the

,y

/

/

station' battery or the alternate

'(f) Verify the cell-to-cell and shutdown battery is inoperable, no terminal connections are clean, actions shall be taken which result in tight, free of corrosion and a reactivity addition, except cooldown.

coated with anti-corroeien or which might rseule in the primary material for the station battery.

coolant system being drained. The alter-the RDS batteries and the

/

/

nate shutdown battery may be inoperable

/

during refueling or shutdown opera-alternate shutdown battery.

/

/

tions es long as containment int.egrity (g) Verify that the battery charger

/

for the main.steen line is established.

for the station battery and the RDS batteries will suppy at 6

If Specification A.5 is not met a normal orderly shutdown of the reactor shall be least 30 amperes at a minimum of 135 volta for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

147 Ameagment yo, gg, pg, gg, gg, pg,,yy May 31. 1989 iO689-0097-NLO4 -

l

'. ~ \\

I Surveillance Requirements Limiting Conditions for Operation 11.3.5.3 DERGENCY PCMER SOURCES (Contd) 11.4.5.3 DERGENCY POWER SOURCES (Contd)

(h) Verify that the capacity of initiated within one (1) hour and the station battery, the RDS l

the reactor shall be shut down as batteries and the alternate shutdeus l described in Section 1.2.5(a) within battery is adequate to eupply and i

twelve (12) hours and shut down as maintain in OPERABLE status all described in Section 1.2.5(a) and of the actual emergency loada (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

for the design time interval when the battery is subjected to a 7.

One RDS uninterruptible power supply battery rervice tes':.

The design including battery may be out of service time interval for the EDS batteries as described in Section 3.1.5 Action 3.

is one heur, two hours for the station battery and seventy-p 8.

Dstring reactor pwar operation, the

d two hours for the alternate 138 kV line may be out of service for repair for periode up to three (3) days.

shutdown battery.

fH (i) Test and calibrate the 2400 volt y

9.

If Specification A.8 is not met, a bus undervoltage trip control normal orderly shutdown shall be components as follows:

I initiated within one (1) hour and the react:or shall be shut down as (1) The undervoltage relays described in Section 1.2.5(a) within 127-10XY, XI and YI will twelve (12) hours and shut down as drop out on decreasing described in Section 1.2.5(a) and voltage of no lower than (b) within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.*

107.1 volta, after a delay of.< 6 seconds.

B.

During power and refueling operations,

-(2) The auxiliary timing relay I

the 2400 volt bus undervoltage components 162-104 will be actuated l

shall be operable or placed in the tripped af ter a 10 2 0.5 second condition, except during the monthly time delay upon receiving e channel functional testing period.

signal from all thrwe (3) undervoltage relays.

j 4

148

, Amendment No. !@, $1, $$, $f, 9$, jf, 102 e

March 16, 1990 y

l s

TSB0689-0097-NLO4 l

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.n

  • A Limit 1ng Conditions for Operation Surveillance Requirements 11.4.5.3 D0DtCENCY POWER SOURCES (Contd)
2. Monthly -

(a) Test start diesel generator and operate at least the fire pump as a load to 480. V Bus 23 for at least 20 minutes.

(b) Verify that the cell voitage is 22.0 volta and specific gravity

~

ie 21.2 of each cell of the station battery; and verify that' y

the cell voltage is 16.0 volte

j and specific gravity.is 21.2 on each cell of tha RDS batteries; and verify that the cell voltage

/

is 22.1 volta and specific

/'

gravity is 21.2 of each cell of

/

the alternate shutdown battery.

/

(c) Test operate the rod position motor generator set.

(d) Performa 'a channel functional.

test of the 2400 volt bus undervoltage trip system.

3. W okly -

(a) Verify tho' electrolyte level of

/

each RDS battery pilot cell, the

/

station' battery pilot cell. and the' l'

alternate shutdown battery pilot

/-

l cell is between the minimum and

-maximum. level indication marks.

h-148a Amendment No.19. 42. Jf. 97 d

May 31, 1989 s

fl 1689-0097-MLO4 i

I

m 6

~.

Limiting Conditions for Operation Surveillance Requirements j

11.4.5.3 EMERCENCY PC'.TER SOURCES (Contd)

(b) Verify the pilot cell specific gravity for ES. station and alternate shutdown batteries corrected to 77*F. is 21.2.

(c) Verify the station battery pilot f

cell voltage is 22.0 volts.

[

The US battery pilot cell' voltage is 26.0 volts and the.

]',

alternate shutdown battery pilot cell voltage 22.1 volts.

ox (d) Verify the overall battery voltage Mi

' is 2125 volts for the station-74 j battery the RDS batteries and

+-

the alternate Shutdown bettery.

t (e) Test start the diesel generator l

and run for warm-up period.

t i

(f) Verify that the diesel generator battery electrolyte level above

, elates and overall battery l

voltage is 224 volts.

4. Quarterly - Verify the following:

[

l

\\

(a) That the specific gravity of the diesel generator bettery is appropriate for continued service.

(

9; 148b v.

Amendment No. IG. (f. 5f 97. 102 o

March 16. 1990 TSB0689-0097-NLO4

  • f

- =

I Surveillance Requiremente Limiting Condicions for Operation 11.4.5.3 DERGENCY PCATR SOURCES (Contd)

(b) That the diesel generator battery and battery rack show no visual indication of physical damage or abnormal deterioration 1 and l

(c) That the diesel generator battery terminal connections are clean, tight, free of corrosion and coated with anticorrosion material.

5

5. Sixty months - At least once per Q

60 months during shutdown, verify that the RDS batteries, the alternate y

shutdown bsttery and the station

/

battery capacity is at least 80% of

/

=

the manufacturer's rating when subjected to a performance discharge test. This performance discharge test shall be performed subsequent to the satisfactory completion of the required battery service test of Part 11.4.5.3.A.I.(h).

l l

l

,e l 1

148e Amendment No. 42. 97. 102 Ef l

March 16. 1990 ci oi M]

och 589-0097-NLO4 1

t

v, e *.,=

Bases:

Normal station power can be provided by the station turbine generator, the 138 kV transmission line or the 46 kV line. These sources are adequate to provide emergency a-c power. When none of these sources is available, a single emergency diesel generator rated at 200 kW starts automatically to provide emergency a-c power to 480 V Bus 2B.

The weekly starting test is based on Manufacturer's Bulletin 33743-1 for relubrication protection of moving parts. Diesel generator initiation and output circuit breaker clost.re is accomplished by two voltage sensors: One to detect loss of normel power on Bus 2B; and, the other to provide assurance of generator output prior to automatic closure of the generator output breakar. Additional breaker interlocks are provided to assure that the normal Buses IA and 2A are isolated prior to closing the generator output breaker. This prevents overloading of the generator during the switching period. An undervoltage trip at 589.25% of normal I>

voltage isolates the 2400 volt bus prior to any postulated equipment degradation.

3

>g The operability of the diesel battery and charger is verified by the weekly starting test of the h.

diesel and by the weekly verification of the electrolyte level and overall battery voltage.

g The diesel fuel oil tank is sized for ten-day full-load operation. Three-day supply is considered I

adequate to provide fuel makeup.

A single station battery supplies power for normal station services and is sized for emergency y

uses including valves and controle of Lose of Coolant Accidents. The battery can be charged from the emergency diesel generator output if normal station power sources are not available.

The primary core spray valves and the primary containment spray valve are operated and controlled by power from the station battery. The bcckup core spray valves and backup containment spray valve are operated by power from normal station power sources or the emergency diesel generator.

RDS uninterruptible power supplies (UPS) A, B, C, and D (each consisting of a battery,' battery charger and an inverter) supply each division (except division 5) with electrical power. Each-UPS has outputs of 120 V a-c, 60 Hz, and 125 V d-c.

One of these batterina supplies control power for the emergency diesel generator. Divisions 1 & 2 and 3 & 4 normally receive power from 480 V a-c Buses IA and 2A, respectively. In the event of loss of power to either or both buses.

provision is included for supplying input power from 480 V a-c Bus 2B'which is tied to the y:y-emergency diesel generator. If all 480 V s-c power is lost, the RDS UPS is capable of

/

a

.a 149 4 Amendment No. 10. 42, 63. 11. 94

~

KCorrected 7/6/89 4Tobruary 15. 1989 TSB0689-0097-NLO4 l

l.

m;j

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g,.

Bases (Contd) sustaining. 'its outputs for one' hour.,, The. Station, Battery has adequate capacity to supply and maintain in an operable status all of..the emergency loads during_a Loss of coolant accident plus an assumed Mas of AC ?ower'.fo( two hotirs'. The station, battery and the four (4) RDS batteries will be considered

~

operable if'they'ar~e essentia'l1[. fully charged.and the battery charger is in service.

e Additionally.

prior to thestartup' following the 197,7. refueling outage. successful completion of service testing and' performance discharge testing within,.each operating cycle and each sixty months, re.spectively.

will further, establish battery reliability.

An alternet,e shutdcwn battery supplies power to the main steam' isolation valve. the emergency

~

/

condenser outlet valves and other alternate shutdown equipment.. The battery is sized so that

/

loss of the charger. does not affect operability of the battery for up to' aix (6) days at a

/

minimum of 25*F (nine. (9) days at a minimum of 40*F).

/

. 3 Q.W

x;.

A-.i '

s 149a 4

Amendment No. 19, gy, gy, 31,'pg, 97

'7 May 31, 1989 IS***89-0097-NLO4 Us

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