ML20038C554
| ML20038C554 | |
| Person / Time | |
|---|---|
| Site: | Shoreham File:Long Island Lighting Company icon.png |
| Issue date: | 12/08/1981 |
| From: | James Smith LONG ISLAND LIGHTING CO. |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| SNRC-642, NUDOCS 8112110214 | |
| Download: ML20038C554 (11) | |
Text
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f SHOREHAM NUCLEAR POWER STATION P.O. BOX 618, NORTH COUNTRY ROAD + WADING RIVER, N.Y.11792 m.u.m w2,
December 8, 1981 SNRC-642
('Q," J N 0
Mr. Harold R.
Denton, Director Office of Nuclear Reactor Regulation
'f U.S.
Nuclear Regulatory Commission k
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washington, D.C.
20555 im/ MClo DfA FI
% s, Shoreham Nuclear Power Station - Unit 1 Docket No. 50-322 h
- D* kI~
Dear Mr. Denton:
Enclosed please find sixty (60) copies of our replies to re-
~
quests made by the Advisory Committee on Reactor Safeguards (ACRS) Full Committee Hearing for Shoreham on October 15, 1981.
These replies address:
1.
Loss of AC/ Loss of DC Power - Describes both the offsite and onsite power systems, the actions for restoring AC power in the event of a loss of the grid, and addresses the condition of total loss of power (AC and DC); and 2.
Station Blackout Analysis - Presents a scenario for maintaining control of Shoreham following a complete loss of both offsite and onsite AC power.
The above address specific ACRS concerns such as flow path con-figuration, environmental limits of required equipment, station batteries capability, and required instrumentation and lighting.
The NRB charter, which was also promised at the ACRS Full Com-mittee Hearing, is presently undergoing change and will be for-warded after revision.
If you require additional information or clarification, please do not hesitate to contact this office.
Very truly yours, L.
Smith Manager Special Projects Shoreham Nuclear Power Station RWG:pg Enclosures cc:
J.
Higgins j7g6f J. McKinley - ACRS
[7 bl 8112110214 811200 FC-893 5 PDR ADOCK 05000322 F
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.s LOSS OF AC/ LOSS OF DC POWER OFFSITE POWER Electric power generated at the Shoreham Station is transformed from generator voltage to a 138 Kv transmission system voltage by the main power transformers.
The main power transformers are-connected via steel transmission poles to the station 138 Kv
' switchyard.
Four 138 Kv overhead lines exit the switchyard on two separate rights-of-way and connect to the 138 KV grid at Holbrook, Brookhaven, and Wildwood substations.
The station 138 Kv switchyard buses are continuously energized and serve as the preferred power source for the station safety loads.
Two physically independent lines are provided to the Plant from the 138 and 69 KV switchyards.
Two system auxiliary trans-formers step the nominal 138 and 69 Kv voltages down to the station 4,160 volt power system buses.
This configuration pro-vides redundant sources of offsite power.
ONSITE POWER SYSTEMS AC Power System In order to guard against the remote possibility of the loss of all electrical power from sources outside the plant, three com-pletely independent standby diesel generators and distribution buses are provided.
These emergency buses are arranged such that any two generators can provide power to all of the loads that are deemed essential for the design basis accident.
This redundancy in the emergency d.iesel generator system assures a high degree of reliability in the power supply to the safety related systems.
There is physical and electrical separation of bus sections, switchgear, interconnections, feeders, load centers, motor control centers, and other system components.
Loads important to plant safety are split and diversified with means provided for rapid identification and isolation of faults.
DC Power System The DC power system consists of the following subsystems, each of which is fully independent of the others:
a.
125 V - DC subsystem (ECCS Div I and non-ECCS) b.
125 V - DC subsystem (ECCS Div II) c.
125 V - DC subsystem (ECCS Div III and non-ECCS)
2 d.
125 V - DC subsystem (non-ECCS - Balance of Plant) e.
125 V - DC subsystem (non-ECCS - Balance of Plant) f.
125 V - DC subsystem (non-ECCS - Communications & Security) g.
-+24 V - DC subsystem (non-ECCS - Radiation & Neutron Monitoring Systems) h.
+24 V - DC subsystem (non-ECCS - Radiation & Neutron Monitoring Systems) 125 V DC subsystems associated with ECCS Division I, II & III do not have any interconnections with each other.
The 24 V DC batteries are connected to provide + 24 V DC power to the radiation and neutron monitoring systems and auxiliary trip devices.
The three 125 V DC subsystems supplying non-essential loads are required to provide control power to the balance-of-plant systems.
One of these 125 V DC subryitems supplies uninterruptible power to the station communication and security systems.
A DC system reliability. study has been performed using very con-servative assumptions.
The reliability of having at least two of the three ECCS DC systemc gperable for any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is 0.999999 or approximately 10-change of failure.
Two of three DC systems are adequate for safe shutdown of the plant, given an accident.
Either Division I or II DC system is required in'the absence of an accident.
The probability of having at least one of these two ECCS DC systems oper9 e for any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period is bl
.999999692 or approximately 3x10-chance of failure.
The probability study makes two conservative assumptions.
One is that no credit is given for repair or replacement of any failed DC component.
The second is that a failure of any of the DC-dis-tribution system components is concluded to fail the entire divi-sional system.
The Action for Restoring Offsite AC Power in thd Event of a Loss of Grid All LILCO fossil fuel fired plants have onsite black start capa-bility.
In addition, LILCO has black start capability gas tur-bines throughout the system.
In case of a complete system shut-down the system power dispatcher would restore the system to nor-mal in as short a time as possible by isolating and re-energizing the transmission' network in a sequence which would vary with pre-valling conditions at the time of the blackout.
Auxiliary power will be restored to Shoreham on a priority basis.
3 Of the eight gas turbine driven generators on the LILCO system with black start capability, any one is capable of supplying the total emergency power requirements of the Shoreham plant.
Restoration of the auxiliary power supply to the Shoreham plant is expected tc be accomplished-by black start of the Holbrook (Holtsville G.T.) station and restoration of the 138 Kv lines between Holbrook and the Shoreham 138 Kv switchyard.
This action will restore one of the preferred sources to the Shoreham plant in less than 30 minutes.
As discussed above, the Shoreham station is powered from re-dundant offsite sources and redundant onsite diesel generators and redundant onsite DC systems.
There is no single event and single failure which could mechanistically result in the total loss of power to the station.
Nevertheless, the purpose of this evaluation is to consider the consequences of such a total loss of power event, and to identify the recovery procedure.
The total loss of power (AC and DC) is assumed at a time when the plant is operacing at 100% and experiences no other transient or accident except that caused by the loss of power.
At that time,'all make-up capability to the reactor is lost and SRVs will function automatically to maintain RPV pressure.
As this con-dition continues, the vessel level will begin to drop.
The level at the top of the active fuel will be reached in 20 minutes, the middle of the active fuel in 30 minuter, and the peak clad tem-perature of 2200 F will be reached in about 40 minutes.
Note that actual fuel damage is not assumed to begin until this point in time.
The recovery procedure to restore power is mindful of this time limitation and is discussed in detail below.
In the event of a total station blackout, the operator will manu-ally operate the appropriate 4.16 KV circuit breakers such that upon restoration of power from Holbrook, voltage will be immedi-ately available at the Division 2 emergency buses.
To facilitate.
access to the required distribution equipment, permanently in-stalled 8-hr. battery pack emergency lights are provided between and at the required areas.
The emergency lights are independent of any plant buses.
A viable response scenario to restore AC and DC power is as follows:
- 1. AT THE MAIN CONTROL BOARD, operate the control switches for the normal and reserve supply breakers to the normal 4.16 KV buses to the " PULL TO LOCK" position.
The breakers are listed in Para. 2.
Place the control switches for the normal and reserve supply breakers to emergency 4.16 KV bus 103 in the " PULL TO LOCK" position.
- 2. AT THE NORMAL 4.16 KV BUSES, trip the following breakers by depressing the manual trip latch on the circuit breaker:
v:
5*
_4__
BUS-BKR BKR ID.
BPEIGER DESCRIPTIGI i
IA-3 410 KSS Trans 1Rll-T-003f 1A-4 420 RSS Trans IRil-T-004 1B-2 430 NSS Trans 1Rll-T-003 1B-1 400 RSS Trans 1Rll-T-004 11-11 450 NSS Trans 1Ril-T-003 11-1 460 RSS Trans 1Rll-T-004 12-1 470 NSS Trans 1R11-T-003 12-11 440 RSS Trans IRll-T-004 3.
AT DERGCCY 4.16 K7 BUS 102 a.
Trip the follcwing breakers b/ depressing the trip latch on the circuit breaker:
BUS-BKR BREM'ER DESCPIPTIC:T_
102-3 1P41*P-003B Serv W PP 102-5 lE21*P-013B Core Spray PP 102-6 lEll*P-014B MIR PP 102-7 1Cll-P-017B CRD Etr PP 102-9 IM50*kC-003B RESVS & CMC W Chiller 102-2 3R11-T-004 PSS Trans 102-8 1R43*G-102 Emr D.G.
b.
Close the nor al supply breaker No. 455 (102-1) by depressing the close latch en the circuit breaker.
4.
AT DERGDCY 4.16 KV EJS 103, trip the following breakers by depressing the trip latch on the circuit breaker:
BUS-BKR BKR NO.
BREAKER DESCRIPTION 103-1 435 NSS Trans 1Rll-T-003 103-2 444 RSS Trans 1Rll-T-004 5.
AT EERGECl 4.16 KV EUS 101:
a.
Trip the folicwing breakers by depressing the trip latch on the circuit breaker:
BUS-BKR BREAKER DESCRIPTION 101-1 1Rll-T-003 NSS Trans 101-3 1P41*P-003A Serv W PP 101-5 1Cll-P-017A CT Wtr PP 101-6 lE21*P-013A Core Epray PP 101-7 lEll*P-014A MIR PP 101-8 1R43*G-101 Emer Dies Gen 101-9 3M50*FC-003A RB SVS & CRAC W Chiller b.
Close the reserve supply breaker No. 424 (101-2) bj depressing the c1cse latch on the circuit breaker.
t
-J
g.-,
, 6.
'Ihe plant AC buscs are roa alicined to safely provide power to bus 102 upon restoration of offs 2.te gn.ur frcn I!albrcok. Idli-tionally, Bus 101 is aligned with the reserve sucply, should power restoration come frcm reserve sources.
Indication of emergency bus voltage is displayed on soltmeters located on the thin Control Doard.
7.
Ibstoration of Division 2,125 V DC pc'..tr is from rectified AC through normally closed circuit breakers at IC 1R24*'rC1125 and within battery charger 3R42*EC-Bl.
Similarly, Division 1, 125 V DC pcreer is through battery charger 1R42*EC-Al fed frcra FCC 1R24*:CC1115.
8.
Start necessarf Division 2 loads, or Division 1 loads, as applicable, to ensure safe shutdcwn of tF.c reactor.
'Ihc IIPCI System will first be used to estchlish vessel level control prior to proceeding to an orderly shutdoen.
t e
G S
STATION BLACKOUT ANALYSIS The loss of off-site AC power and the three onsite independent 3500 kW diesel generators used to supply emergency power to the station constitutes the term " Station Blackout".
This scenario is not part of the design basis for Shoreham.
No single event together with a single failure would result in a Station Blackout.
The scenario presented below identifies those actions available to the operator to maintain control of the station while AC power is being restored.
The equipment available following this event is identified in Table 1.
A course of action has been identified which will re-sult in maintaining core and containment integrity for approxi-mately 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the initiating event.
At that time, the reactor will be depressurized and the containment-will be at or below its design pressure and temperature limits.
There are three large sources of water; the Condensate Storage Tank (capacity 550,000 gallons), the Suppression Pool (574,794-gallons at low level), and 2 fire pump storage tanks (capacity 300,000 gallons each).
There are three non-electrically powered pumps; the High Pressure Coolant Injection Pump (HPCI, capacity of 4250 gpm), the Reactor Core Isolation Cooling Pump (RCIC, capacity of 400 gpm), and the fire pump (capacity of 2500 gpm).
The fire pump is diesel driven while the HPCI and RCIC pumps are steam turbine driven.
The flow rates for the HPCI and RCIC pumps are for either low or high pressure conditions in the reactor.
Immediately following a station blackout, the reactor vessel will be isolated and the SRVs will operate automatically to limit RPV pressure.
The resulting drop in RPV level (due to loss of feedwater) will cause both HPCI and RCIC to actuate at the level 2 setpoint.
Both systems will operate to raise the level to the high level trip in a manner of minutes, at which I
time HPCI and RCIC will both trip.
Continued decay heat generation and SRV operation will result in a decreasing water level at which time RCIC is assumed to be manually initiated and throttled to control level.
The HPCI system will not re-start since RCIC has adequate capacity for level control.
The plant would be in a stable condition at this time, requiring minimal operator action.
However, the continued SRV discharges to the suppression pool combined with the RCIC turbine exhaust flow will cause a con-tinued temperature rise of the suppression pool inventory.
Continued SRV discharges and RCIC operation would be feasible until the pool reaches 200 F with RCIC taking suction from the condensate storage tank.
This temperature would be reached after approximately 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> at which time pool cooling must be initiated.
If AC power has been restored, the RHR system would be operated in the pool cooling mode and the plant could proceed to a cold shutdown condition.
However, if it is assumed that AC l
2 power remains unavailable, an alternate cooling path could be established by manually operating the RHR heat exchangers in the steam condensing mode.
The diesel driven fire pump is capable of delivering 600,000 gallons of fire protection water at 3,000 gpm to the RHR exchangor tubes via a local hose con-nection to the service water system.
This connection must be added as discussed below.
All valves required for system operation can be operated manually.
However, it will be necessary to add certain addi-tional instruments to permit more refined control and to more efficiently utilize the remaining sources of water.
This in-strumentation consists of a local level indicator for RHR ex-changer level control, a local pressure indicator for operation of the steam supply valve and a local temperature indicator to enable adjustment of the tube side fire protection water flow, if necessary.
Shoreham is proceeding to add these instruments.
In conjunction with the condensing mode, the RCIC system suction will be transferred to the suppression pool for a better control of suppression pool inventory.
The plant can remain in this mode as long as necessary subject to the limit of the fire protection water supply.
There are numerous sources of additional water, including salt water if necessary, which could be transferred to the fire protection water tanks.
Using these additional water sources, the plant can remain stable using RCIC to control RPV level as required.
Further, as the suppression pool temper-ature increases, the RCIC suction could be transferred back to the condensate storage tank, if required.
The analysis of the station blackout event also considered any possible design or operational limitations which could affect the scenario for plant shutdown described above.
The principal areas examined were flow path configuration, station batteries capability, environmental qualification limits of affected equip-ment, required instrumentation (as described above), and required fire protection water connection.
To establish the flow path configuration, each required component in the HPCI and RCIC systems and the alternate cooling path utilizing the RHR heat exchanger was specifically examined.
It was confirmed that each component required to operate was sup-plied by available DC power or could be readily positioned manually.
With respect to station batteries capability, it was determined that adequate capacity exists to power the required equipment for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
In order to achieve this capability, all unnecessary loads must be tripped within the first hour including the computer inverter.
3 The temperature rise due to loss of HVAC was evaluated in all safety related areas requiring environmental control.
The spaces examined include the control room, computer room, HVAC room el 63-0, chiller room el 63-0, emergency switchgear rooms, battery rooms, diesel generator rooms, reactor building MCC and MG cubicles, the HPCI/RCIC pump turbine area el 8-0, and the service water pump house.
In summary, it was found that the temperature I
will not exceed limits governed by environmental equipment quali-I fication in any area for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
On loss of AC power, the majority of heat gains in these areas (the electrical equipment loads) reduce to zero.
Only DC equip-ment heat gains remain and, as described previously, most of this equipment is shut down to conserve battery power.
Therefore, only j
minimal loads remain and the resultant temperature rise does not i
exceed environmental qualification limits of any safety related l
equipment during this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period.
Accordingly, there is no adverse ef fect on plant operation due to temperature rise during the station blackout event or following that period upon restora-tion of power.
Also considered was the possible isolation of steam supply lines to the HPCI/RCIC turbines due to temperature rise in the reactor building caused by absence of cooling.
The temperature sensors provided for steam leak detection / isolation are set sufficiently above ambient so as to preclude their operation during events other.than HPCI or RCIC steam line breaks.
Continued RCIC (or HPCI) operation would not be limited during periods of loss of area cooling.
The alternate cooling path utilizing the RHR heat, exchanger in the steam condensing mode takes credit for the use of fire protection water.
The present design does not have an interface between fire protection system and the service water system.
Accordingly, a s
local manual connection is being added.
It is designed to allow adequate flow to assure efficient operation of the RHR heat ex-changer.
In conclusion, vital plant functions can be performed for at least 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, without significant degradation, until power is restored follcwing a station blackout event.
___m
. ~..
' TABLE 1 DESIGN BASIS LOADS l
SAFETY RELATED 125V BATTERIES I.
Battery A-lR42*BA-Al A.
Safety Related Loads (Division.I)
Motor Operated Valves for:
Reactor Core Isolation Cooling (RCIC)
Pump Motors For:
RCIC Condenser Vacuum Pump RCIC Condenser Condensate Pump Diesel Generator 101 Fuel Oil Pump Control Power for:
Backup Scram Trip System A Reactor High Level Trip System C High Pressure Core Injection System (Backup Iso-lation Controls)
Reactor Core Isolation Cooling System Nuclear Steam Supply Shutoff System - Division I i
Automatic Depressurization System - Division I Residual Heat Removal System - Division I Core Spray System - Division I Steam Leak-Detection System - Division I Recirculation Pump Trip System - Division I 4160V and 480V Emergency Switchgear - Division I Diesel Generator 101 Safety Related Ventilation Systems - Division I Aux Relay Panel - Division I CO Detection Panel - Relay Room - Division I 2
Field Flashing for Diesel Generator 101 Distribution Panel 1R42*PNLA5 Miscellaneous Division I Loads in Reactor Bldg.
B.
Non-Safety Related Loads Inverter for Computer Power II.
Battery B-lR42*BA-B1 A.
Safety Related Loads (Division II)
Motor Operated Valves for:
High Pressure Core Injection (HPCI)
Main Steam Line Drain System j
Residual Heat Removal System (RHR)
Reactor Water Cleanup System (RWCU)
Pump Motors for:
Diesel Generator 102 Fuel Oil Pump HPCI Condenser Vacuum Pump HPCI Turbine Bearing Oil Pump 1 of 2
. - -. - -. - -, - -. _ - - =
TABLE 1 (CONT'D)
HPCI Condenser Condensate Pump Control Power for:
Backup Scram Trip System B Reactor High Level Trip System B Core Spray System - Division II High Pressure Core Injection System Reactor Core Isolation Cooling System (Backup Isolation Controls)
Residual Heat Removal System - Division II Nuclear Steam Supply Shutoff System - Division II Steam Leak Detection System - Division II Automatic Depressurization System - Division II Recirculation Pump Trip System - Division II Safety Related Ventilatien Systems - Division II 4160V and 480V Emergency Switchgear - Division II Diesel Generator 102 CO Detection - Relay Room -' Division II 2
Field Flashing for Diesel Generator 102 Distribution Panel 1R42*PNLB5 Miscellaneous Division II Loads in Reactor Bldg B.
Non-Safety Related Loads None III.
Battery C-1R42*BA-C1 A.
Safety Related Loads (Division III)
Pump Motor for:
Diesel Generater 103 Fuel Oil Pump Field Flashing for Diesel Generator ~103 Control Power for:
Diesel Generator 103 4160V and 4SOV Emergency Switchgear - Division III Safety Related Ventilation System - Div_sion III CO Detection - Relay Room
' Division III 2
B.
Non-Safety Related Loads Inverter For Uninterruptible Pcwer 2 of 2
.