ML17340A678
| ML17340A678 | |
| Person / Time | |
|---|---|
| Site: | Turkey Point  |
| Issue date: | 02/04/1981 |
| From: | Robert E. Uhrig FLORIDA POWER & LIGHT CO. |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| L-81-37, NUDOCS 8102090347 | |
| Download: ML17340A678 (30) | |
Text
REGULATO INFORMATION D DISTRIBUTION TER (RIBS)
ACCESSION NiR:81020903@?
DOC,C'ATE: 81/02/Oa "NOTARIZED:
NO FACIL Turkey Point Pl ant g
Uni t 3i Florida Power and Light C
50- 'urkey Po)nt Planti Unit Pg Florida Power and Light C
~
.E AUI'HUR A f F ILI ATI QN UHRIG g R,E.
Florida Power 8 Light Co.
REC IP ~ NAhlE RECIPIENT AFF ILIATIOH EISENHUTiD,G ~
Division of Licensing DOCKET 05000250 05000251
SUBJECT:
Forwards re ponse to Encl 2 of NRC 801124 ltr re fire
~ater'upply 8 cable spreading room, Fire brigade size 8 trairing requirements to be implemented 801119.
DISTRIBUTION CODE:
AOOoS COPIES RECEIVEDtLTR ENCL SIZE:
'ITLE:
Fire Protection Information (After Issuance of OP ~ Lic ~ )
NOTFS:
RECIPIENT ID CODE/NAME ACTION:
VARGAi S ~
09 INTERNAL; CHEM ENG BR A
?
IFE Ob OELO NAABACHiT 10 COPIES LTTR ENCL 7
7 5
5 2
1 0
1 1
RECIP IENT ID CODE/NAHE OIR c DI V OF LIC N r PER 02 EG F IL 01 COPIES LTTR ENCL EXTERNAL; ACRS NSIC Ag Ob le 16 1
1 LPDR 03
)~8
February 4, 1981 L-81-37 Office of Nuclear Reactor Regulation Attn:
Hr. Darrell G. Eisenhut, Director Division of Licensing U.
S. Nuclear Regulatory Commission Washington, 0.
C.
20555
Dear Nr. Eisenhut:
Re:
Turkey Point Units 3 5 4 Docket No. 50-250 8 50-251 FIRE PROTECTION This letter responds to the "open" items in Enclosure 2 of your November 24, 1980 letter.
The attachment addresses Fire Mater Supply (Items 3.1.2 and 3.2.3),
Auxiliary Building Corridor (Item 3.2.4),
and Cable Spreading Room (Item 3.2.5).
Me intend to implement the Fire Brigade Size and Training requirements of Section III, paragraphs H and I of Appendix R to 10 CFR 50 published in the Federal Register on November 19, 1980.
We believe that the provisions of the revised Section 10 CFR 50.48 and new Appendix R are adequately satisfied as described in the attachment and as such demonstrate that the public health
'and safety is properly protected.
Very truly yours, 2 f'~~~+
~Robe t E.
Uhoig Vice President anced Systems 5 Technology REU/JEM/mrs Attachments cc.
Mr. James P. O'Reilly, Region II Harold F. Reis, Esquire O o ego eL/,P
RESPONSE
TO STAFF SIMQRY OF REQUIREMENTS TO RESOLVE "OPEN ITEMS" 3.1.2 FIRE WATER SUPPLY 3.2.3 NRC CONCERN:
"In the Fire Protection Safety Evaluation Report, we were concerned about the adequacy of the screen wash pumps as an alternate source of water for fire fighting as well as the ability of the system to provide the water supply for fixed systems.
By letter dated iiay 7, 1980 the licensee proposed to install a spool piece to interconnect the screen wash pump to the fire protection water system.
- However, the licensee did not want the spool piece to be permanently connected, because of the, possibility of contaminating the domestic water supply, which is connected to the water supply.
The licensee also proposed to administratively control the volume of dedi-cated fire protection water in the gravity tanks and the raw water storage tanks.
Because of the high rate of failure, administrative
- controls, by themselves
- alone, are not acceptable.
We rind that the screen wash pumps cannot be considered as an acceptable backup to the fire water supply.
In order to meet the requirements of Section III, Paragraph A, of the proposed Appendix R to 10 CFR 50, the licensee should provide a verti-cal standpipe for the existing 500,.000 gallon raw water storage tank.
The licensee should also install a new water storage tank as a
secondary source of fire water.
An automatic starting diesel fire pump with adecuate capacity is required to supply water to the fire loop from the secondary source."
FPL RESPONSE:
'1 Section 8.5.2 of the fire protection report presents the design basis for the fire evaluations.
The plant areas analyzed were those considered to be related to safe shutdown capability.
Although fire suppression capability is available, no credit was assumed for miti-gating the consequences of a design basis fire, i. e.,
the fire is assumed to burn at i.ts maximum burnout until the available fuel is consumed.
The Fire Hazards Analysis concludes that safe shutdown conditions can be achieved and maintained for the postulate design basis fire without the use of 'the available fire suppression capability.
FPL's report, "Fire Protection A Reevaluation of Existing Plant Design Features and Administrative Controls", submitted 2/25/77.
1
nonetheless, water for fire suppression is available from two primary sources (as stated in Section 6.1 of report.. THe: gravity tank has a desi'gn capability of 100,000 gallons and the raw water storage tank has a design capability of 500,000 gallons.
Section 3.14 of the plant technical speci-fications concerning the Fire Protection System administratively places minimum water volumes to be contained in each of these separate tanks.
The gravity tank (elevated storage tank) is specified to have a minimum of
~
30,000 gallons of water available for fire protection.
The raw water stor-age tank is specified to have a minimum of 150,000 gallons available for fire protection.
The tank capacities and technical specifications illustrate that water is available on site for fire protection.
l An alternate source of water for firefighting shall also be available by way of the screen wash pumps.
The maximum area of water demand was recalculated to determine the adequacy of the aforementioned pumps.
As a result of our
- analyses, we have determined that these pumps do indeed provide an adequate alternate source of water for fire suppression.
The calculations in support of this conclusion were part of attachment 4 to Florida Power and Light Company's letter of i1ay 7, 1980 concerning Fire Protection.
Deluge system flow rates for the hydrogen seal oil unit, the auxiliary transformer and the main transformer were added to our analysis which was submitted to the NRC Staff on Hay 21, 1979.
(See attachment 4 to Florida Power and Light Company's letter of Hay 7, 1980.)
We do not intend to permanently connect the screen wash pumps to the fire water system.
The fire water system is connected to the domestic water supply.
In leakage from the screen wash pumps through a permanent connec-tion to the 'fire water system could contaminate the domestic water supply.
- Thus, a permanent condition is not considered feasible.
A temporary connection has been proposed, in lieu of a permanent hard pipe connection.
Appropriate connec'tions have been installed and a rolled-up section of hose has been located in the area to facilitate rapid connection between the screen wash pumps and fire water system.
If the screen wash pumps were connected to.the fire water supply, appropriate measures would then be taken to prevent and monitor in leakage to the domestic water supply.
The only requirements in Appendix R,Section III. A. that the system des-cribed above does not meet are:
1.)
The requirement for a permanent connection to the fire water supply and, 2.)
The requirement for a vertical standpipe to insure minimum ~ater volume.
The need for a permanent connection to the fire water supply is precluded by designing the screen wash pump to fire main connection for rapid hookup (approximately 5 minutes).
Administrative controls have classically been utilized to ensure minimum water volumes for all safety-related tanks (i. e., Refueling Water Conden-sate Storage, etc.).
Ve therefore contend that administrative controls as established in the Technical Specifications with the appropriate limiting conditions for operation will ensure minimum water volume in the Raw Water Storage Tank and in the gravity tank.
1 FPL's report, "Fire Protection - A Reevaluation of Existing.Plant Design Features and Administrative control," submitted 2/25/77.
Based on the above discussion, the fire water supply system as described above meets the intent of Section III. A. of Appendix R and no further modifications are required.
3.2 AUXILIARYBUILDING CORRIDOR NRC CONCERN:
"In the Fire Protection Safety Evaluation Report, it was our concern that a fire in the auxiliary building corridor could damage redundant safety-related cables that could affect both units from achieving cold shutdown.
By letters dat'ed Hay 7, 1980 and June 9, 1980, the licensee provided additional information.
The licensee stated that all cables in the corridor are coated with a fire retardant coating.
The licensee also stated that although some of these cables are needed for the safe cold shutdown, the units can be placed in hot shutdown
- mode, and these specific cables could be repaired or replaced to bring the units to a cold shutdown condition.
The additional information submitted by the licensee has been previoulsy:.evaluated.'and.
the:,
licensee informed that it does not meet our concerns.
The licensee has not demonstrated that adequate protection features have been provided for cables and equipment of redundant systems important to achieving safe shutdown conditions to ensure that at least one means of achieving such conditions survives postulated fires.
In order to meet the requirements of Section III.G. of prt)posed Appendix R to 10 CFR Part 50, the licensee should provide the following:
1.
An automatic sprinkler system in the auxiliary building corridor to protect the cable trays from an exposure fire; 2.
A fire detection system in the chemistry laboratories (hot and cold labs) and in the new laundry facility; 3.
Three-hour fire doors in the auxiliary building for the chemistry laboratories (hot and cold labs) and new laundry facility; 4.
Three-hour fire rated fire dampers for ventilation ducts in the chemistry laboratories (hot and cold labs) and the new laundry facility; and, 5.
The plastic barrels used to collect radiation 'protective clothing be replaced with barrels made of a fire retardant material.
To meet our fire protection guidelines, alternate shutdown capability should be provided when safe shutdown cannot be ensured by barriers and detection and suppression systems because of the exposure of redundant safe shutdown equip-ment, cabling or components in a single fire area, to an exposure fire, or fire suppression activities, or rupture or inadequate operation of fire suppression systems.
To meet Section III, Paragraph G. of the proposed Appendix R to 10 CFR Part 50, the licensee should provide an alternate shutdown capability independent of this area.
The alternate shutdown system should meet the requirements of Section L, Paragraph III of proposed Appendix R to 10 CFR Part 50."
FPL RESPONSE:
The following modifications have been completed.
1.)
Fire detection in the chemistry laboratories and laundry facility.
2.)
Three-hour fire rated doors for the chemistry laboratories and laundry facility.
3.)
Three-hour fire rated fire dampers for ventilation ducts in the chemistry laboratories and laundry facility.
4.)
Replacement of the plastic barrels used to collect radiation protective clothing with barrels made of steel with automatic closing fusible link lids.
Attached (see Attachment
- 1) are the safe shutdown evaluations for the auxiliary building hallway and the cable spreading room.
Note tfiat these evaluations were performed assuming total loss of all cable functions.
Total loss of all cable functions, however, is not considered credible based on the design basis fire considerations for this area (see Section 8.5.2.20 of "Fire Protection - A Reevaluation of Existing Plant Design Features and Administrative Controls" ) and a number of inherent design features at the facility.
These features include:
1.
Safe shutdown related cables are separated in accordance with criteria licensed and accepted by the NRC.
(Note:
This separation criteria is minimum spacial criteria between trains.
Specific "A" and "B" train functions may be separated far in excess of the minimum criteria.)
2.
All safe shutdown cable trays and trays'n safe shutdown fire areas are sprayed with Flamemastic 71A which has been demonstrated to not only protect cable functions under certain fire conditions, but also to prevent fire propagation under extreme, fire. conditions (see Section 3.1 of our fire protection report and Attachment 2).
3.
The auxiliary building hallway is occupied continuously.
4.
Early warning fire detectors are installed.
5.
Our fire methodology for determining burn rates and gas and surface
temperatures was developed in the fall of 1976.
Florida Power and Light Company's technique and results have been shown to be conservative when compared to similar methodology developed by Sandia.
Even based on our con-servative methodology, design basis fires in areas of large cable concentrations were considerably smaller than any used in Florida Power and Light Company or San-dia test efforts.
- Thus, no cable damage or fire propa-gation will occur as a result of design basis fires in these areas.
6.
Based on our conservative fire methodology and conserva-tive assumptions on suppression capability (i. e.,
50%
effectiveness),
Florida Power and Light Company demonstrated that although the fire could burn till ex-tinguished in each area, the manual suppression capability at the facility is sufficient to extinguish all design basis fires.
Me believe that the items we have implemented in conjunction with the features described
- above, ensure that fire-induced cable function loss cannot occur in the auxiliary building hallway.
Even with the incredible assumption that total loss of cable functions has occurred we have demonstrated the capability to bring the plant to cold shutdown.
3.2.5 CABLE SPREADING ROOIf NRC CONCERN:
"In the Fire Protection Safety Evaluation Report it was our concern that a fire in the cable spreaidng room could prevent both units from achieving safe shutdown.
By letters dated Hay 7, 1980 and June 9, 1980, the licensee provided additional information.
The licensee stated that all cables in the cable spreading room are coated with a fire retardant coating.'he licensee also stated that although redundant safety-related cables and equipment may be damaged by a fire, the units can be placed in a hot shutdown condition and necessary repairs and/or replacement be made to bring the units to a cold shutdown condition.
The additional information submitted by the licensee has been previously evaluated and the licensee informed that it does not meet our concerns.
The licensee has not demonstrated that adequate protection features have been provided for cables and equipment or redundant systems important to achieving safe shutdown con-ditions to ensure that at least one means of achieving such.
conditions survives postulated fires.
In order to me'et the requirements of Section III.G. of Proposed Appendix R
to 10 CPR to Part 50, the licensee should provide the following:
1.
An automatic'ater suppression system in the room; 2.
A booster hose station in the room; and, 3.
Access doors from the cable spreading area to the turbine building to have l-l/2 hour rating.
To meet our fire protection guidelines, alternate shutdown capability should be provided when safe shutdown cannot be ensured by barriers and detection and suppression systems because of the exposure of redundant safe shutdown equipment, cabling or components in a single fire area, to an exposure fire, or fire suppression activities, or rupture or inadequate operation of fire suppression systems.
To meet Section III, Paragraph G. of the proposed Appendix R to 10 CFR Part 50, the licensee should provide an alternate shutdown capability independent of this area.
The alternate shutdown system 'should meet the requirements of Section L, Paragraph III of proposed Appendix R to 10 CFR Part 50."
FPL RESPONSE A low flow capacity nozzle was installed prior to October 30, 1980 in the area immediately outside the cable spreading room.
The length of the hose. at this station is sufficient to reach all areas inside the cable spreading room.
The design and implementation package for 1-1/2 hour fire rated doors between the cable spreading room and the turbine building has been completed.
They are currently scheduled to be installed.
Attached (see Attachment
- 1) are the safe shutdown evaluations for the auxiliary building hallway and the cable spreading room.
As indicated pretiously, these evaluations were preformed assuming total loss of cable functions.
Total loss of all cable functions, however, is not considered credible based on the design basis fire considerations for this area (see Section 8.5.2.41 of our fire protection report) and a number of inherent desi'gn features at the facility.
These features include those identified for the auxiliary building hallway.
In addition, combustibles are no't stored in the cable spreading room and there is no equipment in the cable spread-ing room which requires major sources of combustibles for operation or maintenance.
Me believe that the items we have implemented in conjunction with the features described above ensure that fire-induced cable function loss cannot occur in the cable spreading room.'ven with the incredible assumption that total loss of cable functions has occurred, we have demonstrated the capability to bring the plant to cold shutdown.
ATTACEKNT 1 SAFE SHUTDOWN EVALUATIONS FOR THE AUXILIARYBUILDING HALLWAY AND CABLE SPREADING ROOM (Attachment to FPL ietter of June 9,
1980 Number L-80-174)
of 19 SAFE SHUTDOWN EVALUATIONS FOR THE AUXILIARYBUILDING HALLWAY AND CABLE SPREADING ROOM Recent NRC correspondence has requested a demonstration of safe shutdown assuming loss of'able function in the Auxiliary Building Hallway or the Cable Spreading Room and provide what is called a "task/manpower" evalua-tion to demonstrate that it can be accomplished.
This evaluation was conducted in the following manner:
A.
Determine the time restraints wherein operator action is required based on safe shutdown plant parameters.
B.
Determine if it is possible to control plant parameters assuming loss of cable function.
C.
Determine if the task or control methods of item B and the operator or manpower time restraints of item A are appropriate to assure safe shutdown.
Therefore, if it is estimated that it would take an operator some 4 or 5
minutes to perform a control function and he has,
- say, 8-10 hours to complete it, no further evaluation was required.
Also, current plant emergency procedures were used during this evaluation.
These procedures have been reviewed and approved in accordance with the license and are demonstrated/performed on a regular basis.
Reactivit Control For Maintainin Hot Safe Conditions The reactivity control system assures safe subcritical cote conditions following reactor trip and during all phases of unit heatup and cooldown operations.
The reactivity control system consists of two independent reactivity control subsystems; the rod cluster control (RCC) assemblies and boric acid injection via the charging system.
The RCC assemblies are divid-ed into two categories comprising control and shutdown rod groups.
Following an equilibrium xenon full power trip with the reactor coolant system (RCS) maintained at hot conditions, excess shutdown/sub-critical margins for full RCC insertion at beginning and end of cycle are:
<<BOC POST TRIP EXCESS SHUTDOWiil Unit 3
-4.20%
d, k/k Unit 4,
-3.74% 6 k/k
<<EOC POST TRIP EXCESS SHUTDOWN
-4.07% 6 k/k
-3.56%
8 k/k
":Unit 3 5
4 are for seventh and sixth cycles respectively.
2 of 19 The reactivity worth associated with complete decay of pretrip full power equilibrium Xenon is +2.57 and +2.60% 6 k/k for Unit 3 and 4 respectively.
Therefore, the minimum available subcritical margin for BOC and EOC are
-1.63% 6 k/k and -1.57% 5 k/k for Unit 3 and -1.14% 6 k/k and -0.96% 6 k/k for Unit 4.
Hence post trip subcritical conditions for maintaining, long term RCS hot conditions are assured via full RCC insertion and is independent of off-site power availability.
Boron addition via the charging pumps is available from two independent boric acid solution sources.
The primary boron injection flow path to the charging pumps is established fzom the discharge of one of the boric acid transfer
- pumps, each of which can be aligned to take suction from any of the three boric acid storage tanks containing a.minimum of 20,000 ppm boron solution.
The secondary oz backup boric acid addition source is the charging p ump ump direct gravity feed line from the refueling water storage tank (RUST),
which contains a minimum of 1950 ppm boron solution.
In addition, system provisions exist for aligning either unit's RMST to the suction of either unit's chargi'ng pumps by equalizing heads in the tanks using the SI suction header.
Two independent boration paths are available from the discharge of any one of the three available charging pumps.
Plow to the loop "A" cold leg and/or the loop "C" hot leg is available in addition to charging flow to the reactor a
coolant pump seals.
As indicated above, the boron addition portion of the reactivity control system is not required in order to maintain the unit in the hot condition and its function, therefore, is not required following loss of off-site power.
Steam Generator Heat Removal For Haintainin Hot Safe Conditions Following r'eactor trip or a unit runback, steam generator heat removal is accomplished by the condensate/main feedwater system throttled back to the steam generator no load condition with steam relief to the condenser via four turbine bypass valves.
One tzain, i.e.,
one condensate pump and one main feedwater pump is sufficient to meet steam generator heat removal requirements.
Condenser hotwell makeup can be obtained from the condensate storage tank which is supplied from the-water treatment plant.
Steam generator heat removal -under hot RCS conditions can be accommodated for a prolonged period ox time
~
3 of 19 In che evenc that turbine bypass, condenser vacuum or condensate/feedwacer pump capabilities are unavailable, steam generator heat removal is main-tained by the auxiliary feedwater system with steam relief via the three atmospheric dump va1.ves (located upstream of their respective main steam
-isolation valves).
Auxiliary feedwater is provided by three turbine driven auxiliary feedwater
- pumps, each with a nominal pumping capacity of 600 gpm.
Any single turbine driven pump is capable of supplying the post trip require-ments of either unit.
One turbine'riven pump is also capable of supplying the auxiliary feedwater requirements of both units 19.2 minutes following full power trip of both units.
Steam supply to th'e auxiliary feedwater pump turbines is provided via motor operated valves on each of the Unit 3 or 4 steam generators upstream of their respective main steam isolation valves (iSIV's).
Steam flow from any one oi the six steam generator supply valves is sufficient to provide rated auxiliary feedwater pump flow for boch units.
In addition, a manual cross tie to the Units 3 and 4 main steam header (downstream of the Nsxv's) and a manual cross tie to the fossil Unit 1 and 2 desuperheated steam header are available to supply the auxiliary feedwater pump turbines.
The 'uxiliary feedwater pumps are normally aligned co take suction from the Unit 3 and 4 condensate stora'ge tanks and discharge to the Unit 3 and 4
If additional Unit 3 or Unit 4 condensate storage tank inventory is desired, it can be provided from the water treatment plane or the primary water storage tank.
The availability of makeup allows the unit(s) to be maintained at hot conditions for a prolonged period of time.
In the event chat loss of off-site power occurs with the unit(s) at hot shutdown or the unit(s) trips because of loss of off-site power, all auxiliary steam supply valves and pump discharge valves operate automatically and receive their control and power form the DC and vital AC buses.
A single failure of either DC or vital AC bus will not prevent the auxiliary feedwacer system from meeting its design flow requirements.
Following a loss of off-site power the water inventory from the condensate storage
'canks is the only immediately available source to the auxiliary feedwater system.
Each condensate storage tank has a design capacity of 250,000 gallons, The minimum allowable inventory in a condensate storage 10
4of 19 tank associated with an operating unit is 1S5,000 gallons, which is sufficient for 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> at hot conditions and subsequent cooldown to the residual heat removal operating, window, or 23 hours2.662037e-4 days <br />0.00639 hours <br />3.80291e-5 weeks <br />8.7515e-6 months <br /> at hot RCS conditions.
Approximately 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> after the loss of off-site power the water treatment system can be loaded on the vital bus as per procedural steps provided in the "Blackout Operation" Emergency Procedure Eor Units 3 and 4.
This will allow up to 200 gpm makeup to each condensate storage tank whereas only 125 gpm per unit is required to maintain steam generator level after 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />.
Operation in this manner will allow the unit{s) to be maintained in the "blackout" hot condition until off-site ~power is restored.
In addition, a
primary water pump may be loaded on the vital bus, which would allow the unit(s) primary water storage tank (a design capacity of 160,000 gallons each) to be pumped to either or both condensate storage tanks.
Pressurizer Heater Control for ~aintainin Hot Safe Conditions Following reactor trip pressurizer heater capability is required in order to maintain the unit at approximately 2200 psig pressure, which is normally associated with hot RCS conditions.
Loss of total pressurizer heater function is not considered a likely event since the two heater backup groups are powered by train A and train B 480 V load. centers.
While it is preferred to have pressurizer heater capability to maintain RCS pressure in the post trip condition, it is not a safety requirement nor is it required following loss of off-site power.
RCS pressure can be allowed to drift slowly downward until diesel generator loads are sufficient-ly low to allow the "A" backup group to be energized.
(If both diesel generators have started and are operating properly, sufficient load capability exists to reload the backup group immediately).
If it were postulated that pressurizer heater function cannot be reestablished, this would not preclude RCS cooldown to RHR levels following the addition of adequate boron solution to accommodate RCS cooldown.
Component Coolin for Miintainin Hot Safe Conditions The component cooling water (CCW) system has three 100%
pumps and three 50%
heat exchangers and is normally operated with one pump and two CCW heat exchangers in service during full power operation.
Two of the pumps are powered by train "A" and one by train "B" 4160 V switchgear.
In addition,
5 of 19 the re is a normally isolated S inch crossover line between the Unit 3 and 4 component cooling water systems.
This cross tie between units can provide sufficient heat removal capability to maintain a unit at hot shutdown until its normal CCM flow is reestablished.
Following a loss of off-site power at least one CQC pump and one intake cooling water pump per unit will be loaded on the diesels, which assures required CQP flow to post trip compo-nenes.
Availability of the 8 inch CCW cross tie between units is not affected by loss of off-site power.
Electrical Power Su 1
For Haintainin Hot Safe Conditions Following reactor trip, normal off-site power is provided by the Unit 3 and 4 start up transformers via the 240 kv switchyard.
The loss of power to the Unit 3 startup transformer results in the loss of the Unit 3 "B" train 4160 V switchgear only since the "A" train has an alternate feed from the Unit 4 startup transformer, which could be made available to service Unit 3 "A" train requirements.
Under these conditions both diesel generators would start.
The diesels can supply the vital portions of the Unit 3 "A" and/or "B" trains. If the Unit 4 startup transformer is lost, the consequences are the same.
Hence a loss of one of the startup transformers would not preclude maintaining both units at hot conditions for a prolonged period of time.
A total loss of off-site power caused by grid upset or loss of transmission lines could result in interruption of off-site power to Turkey Point station.
Transmission/grid difficulties of this nature occurring coincident with an event such as a design basis fire is unlikely.
- However, assuming total loss of off-site powerTurkey Point Unit 3 and 4 diesel generators would start and pick up train "A" and "B" vital loads for both units.
Units 3 and 4
are also designed to accommodate the starting failure of one diesel genezator and still maintain safe hot conditions.
As indicated earlier the water treatment system can be loaded onto the vital bus which will allow Units 3 and 4 to be maintained at the "blackout" hot condition until off-site power is restored to the 240 kv switchyard.
In addition to the above electrical provisions and those normally followed by the system load dispatchers, there are two other on-site methods available for restoring power to the 240 kv switchyard.
The Turkey Point 1 and 2 oil fired units have the design capability for black startup following 12
6 of 19 a total loss of off-site power.
As required, fossil Unit 1 and 2 2500 KN "Black Start" diesels (five available) can be started following the loss of off-site power.
Unit 1 and 2 auxiliary loads are carried by the black start diesels as required to support sequential restart of the units..
Following Unit 1 and/or 2 startup, the associated main -generator can be aligned to restore power to the nuclear Unit 3 and 4 240 kv switchyard.
The second on-site method for restoring power to the Unit 3 and 4 240 kv switchyard is conducted using the five Unit 1 and 2 black start diesels, which can accept load in one minute in the dead-line mode and in ten minutes in the normal or peaking mode.
The black start diesels can be alignd to energize the Unit 1 and 2 240 kv switchyard via the Unit 1 and 2 startup transforaer.
(The Unit 1 and 2 switchyard is normally aligned to the Unit 3 and 4 switchyard.)
Thus plant operations personnel can make available to the nuclear units fossil unit power sources.
13
7 of 19 Reactivit Control Function Re uired for Cold Conditions As indicated earlier, sufficient subcritical margin is obtained for both BOC and EOC core conditions from the full RCC portion of the reactivity control system to maintain the unit(s) in the hot condition indefinitely.
Prior to initiating cooldown, the RCS is borated to a boron concentration level that assures sufficient subcritical margin when the unit ultimately reaches cold conditions.
Conservative values of Unit 3 and 4 post trip reactor coolant boron concentra-tions for BOC and EOC should be around 810 ppm and 50 ppm respectively.
Conservative values for the required RCS boron concentrations to assure sufficient subcritical margin at cold RCS conditions would be approximately 1162 ppm and 691 ppm at BOC and EOC respectively.
Boration of the reactor coolant system (as discussed earlier), prior to RCS cooldown is provided by the charging system in conjunction with letdown to the chemical volume and control system (CVCS) in order to maintain pressurizer level.
Normal letdown flow from the reactor coolant system is 60 gpm. It is routed through the regenerative heat exchanger, one letdown orifice and through the non-regenerative heat exchanger to the volume control tank (VCT) or holdup tanks.
Maximum letdown capability via this path is 120 gpm and is achieved by placing another letdown orifice into service.
If CCW is out of service to the non-regenerative heat exchanger letdown is diverted directly to the VCT or if there is a high level in the VCT, flow is diverted to the noldup tanks.
Charging flow to the regenerative heat exchanger is balanced with RCS letdown flow in order to achieve the proper temperature drop prior to going through the letdown orifice(s). If letdown via this path were required under emergency conditions and charging flow was not available to the regenerative heat exchanger, the letdown line out-side of containment could be isolated causing letdown flow to be relieved to the pressurizer relief tank via the letdown line safety valve.
A second normally available letdown path of 9 gpm is from the reactor coolant pump seal water return line to the VCT via the seal water heat exchanger.
An additional 15 gpm excess letdown line is also available for letdown flow, which is directed through the excess letdown heat exchanger to the seal water return line. If CCM is unavailable to the seal water and excess letdown heat
8 of 19 ezchanger, ezcess letdown could be diverted to the reactor coolant drain tank and seal water. return or bypass flow could be relieved to the pressurizer relief tank via the seal water return line safety valve.
All letdown line flow paths require instrument air for the required valving operations.
However, since the unit(s) can be maintained in a safe hot condition for extended periods, sufficient time ezists for reestablishing the instrument air system or to execute manual operations to restore necessary letdown line flow.
The compressor(s) may be loaded on the diesels.
The times required to borate the RCS prior to initiating cooldown have been analyzed the various charging and letdown path combinations:
BOC CONDITIONS EOC CONDITIONS 2 BTP's, 2 CP's 2 Orifices 1 BTP, 1
CP, 1 Orifice 1BTP, 1CP Balanced for ex-cess letdown and RCP sealwater re turn 2 CP's from E4ST, 2 Orifices to PRT 1
CP from EST 1 Orifice to PRT 1
CP from RNST Balanced for ex-cess letdown and sealwater return
- 0. 13 0.26 0.64 2.6 5.1
- 12. 8
- 0. 23 0.45 2.9 5.7 14.2 Key:
= charging pump BTP ~ boric acid transfer pump RUST = refueling water storage tank PRT - pressurizer relief tank Since long term hot conditions can be maintained for either unit, sufficient time is available for even the worst case boration conditions.
The operator has an alternate means to borate prior to initiation of cool-down.
He may elect to maintain hot RCS temperature and allow RCS pressure to fall to approximately 1300 psig and then inject the 20,000 ppm boron
9 of l9 solution contents of the boron injection tank (BIT) using the safety injection system.
For this method of boration, RCS pressure would normally be reduced by turning off the pressurizer heaters.
He also has the option of allowing the pressurizer to vent to the pressurizer relief tank.
All equipment necessary to achieve the above mentioned boration combinations can be powered by the unit vital buses thus, their availability is not af-fected by loss of off-site power.
Hoderator Shrink Hake-U Re uired Durin Cooldown Following proper boration, of the reactor coolant system to required cold shutdown concentrations, the charging system also provides reactor coolant system makeup to accomodate RCS shrink during cooldown.
Two independent sources of makeup water are available to the suction of the three charging pumps.
The normally used water source is obtained from the primary water storage tank via the primary water pumps.
The second independent source is the RWST gravity feed line. If the unit were in the station blackout condition, the makeup source from the RWST would be used since the primary water system pumps are powered from a non-vital power source.
A third source of RCS makeup water for both units can be made available from the spent fuel pools via the spent fuel pool cooling pumps.
Spent fuel cooling pump discharge can be aligned to the CVCS letdown return line upstream of the volume control tank divert valve or to the downstream side of the reactor coolant filter.
Either of these two flow paths can provide makeup water to the volume control tank.
With a spent fuel pool capacity of approximately 7800 gal/ft, less than 4 feet of spent fuel pool inventory from a pool is required to accommodate moderator shrinkage for its unit.
This minor reduction in spent fuel pool inventory will have no'-appreciable affect on spent fuel cooling or shielding requirements.
If it were postulated that none of the three charging pumps are available, safety injection (SI) via the boron injection tank (BIT) can be used to provide precooldown boration and reactor makeup.
As discussed
- above, the RCS pressure would be reduced to approximately 1300 psig and SI flow through the BIT would provide adequate cold shutdown concentration boration.
Following proper boration, RCS cooldown using secondary steam generator heat removal is initiated with the SI system providing RCS makeup as needed to accommodate RCS shrinkage.
The SI pump(s) is powered by a vital bus(es) and 16
10 of 19 its operation therefore is unaffected by loss of off-site power.
Steam Generator Heat Removal For Unit Cooldown As discussed earlier, steam generator heat removal during RCS cooldown will be conducted with the condensate/main feedwater system.
In the event of condensate/main feedwater system malfunction or if off-site power is not available, steam generator heat removal during RCS cooldown will be provided by the auxiliary feedwater system with steam relief via the atmospheric dump valves.
Steam generator heat removal is secured following servicing of the residual heat removal system.
Component Coolin Re uirements For Cold Conditions The heat load on the CCW and intake cooling water (ICW) systems during RCS cooldown is approximately the same as that required to maintain hot condi-tions Thus the minimum required components and CCW unit cross tie capabili-ties discussed earlier are also applicable during this phase of RCS cooldown.
Residual Heat Removal (RHR) For Maintainin Cold Conditions Following RCS cooldown and depressurization using steam generator heat
- removal, the residual heat removal system can be placed in service at 350 F
with RCS pressure at or below 465 psig (normally placed in service about 450 psig, system designed for 600 psig).
RHR heat removal capability is dependent upon available
The following equipment combinations and heat removal capabilities are sufficient to match core decay heat within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> after shutdown.
2 ICWP's, 2
CCWP's 3 CCKX's, 2 RHRP's 1 ICWP, 1
CCWP 2
CCWHX's, 2 RHRP's 1
- ICWP, 1 CC!N 2
CCWHX's, 1
RHRP HEAT RE~ifOVAL RATE (Btu/hr) 3.'4 x 10 8 2.1 x 10 8 1.8 x 10 8 KEY:
ICWP's CCWP's CCHX' RHRHX' RHRP '
ICW CCW CCW RHR RHR pumps pumps heat exchangers heat exchangers pumps 17
ll of 19 All equipment necessary to achieve the above RHR heat removal combinations is powered bv the unit vital buses and is therefore available under loss of off-site power conditions.
After placing the RHR system into service, the op-erator can elect to continue the RCS cooldown'o cold refueling conditions or maintain the RCS at any desired temperature below 350'F.
Auxiliary Buildin Hallway The normal shutdown cable functions in the Auxiliary Building Hallway are shown in Table 1.
Loss of these cable functions would require various off-normal operator actions in order to maintain safe plant conditions.
Sufficient control room personnel are immediately available to perform the required hot shutdown and boration actions in a timely manner.
A summary of necessary actions follows:
Maintain Hot Shutdown Conditions A.
Dispatch one member of control room staff to the Unit 3 Load Center Room and place Unit 3 charging pump switches in remote.
B.
Dispatch a second member of control room staff to the Unit 3 C.
D.
charging pump room, via outside entrance from Boric Acid Storage Tank Room, and establish manual charging pump control.
From control room, verify that a sufficient number of auxiliary feedwater steam supply valves are open.
From the control room, start an additional Unit 4 component cooling water pump and verify Unit 3 component coo'ling water pump operation.
F.
Use the 4C steam generator level to infer and maintain level in the 4A and 4B steam generators.
18
12 of 19 Boration of Reactor Coolant System A.
Zn the event that all four Boric acid transfer pumps are in operable, initiate boration using the refueling water storage tank source as discussed earlier.
As already indicated, the units can be maintained in the hot shutdown mode for days if desired.
This would allow sufficient time,to repair some or most ot the damaged circuits listed in Table l.
At a minimum; (1) one control room level channel from the 4A and 4B steam generators should be repaired or duplicated outside the control room and (2) residuel heat removal pump control circuits should be repaired or bypassed at the 4160V switchgear before cooldown of the reactor coolant system begins.
Reactor coolant system cooldown should proceed using currently established plant procedures with the assistance of Table l.
As procedures call for the positioning of valves that may not be operable, control room personnel can be disoatched to make the valve adjustments manualLy.
Valve operation in support of entering into the residual heat removal mode would be conducted in the same manner.
Once the units are on residual heat removal cooling, reactor coolant system temperatures can be maintained at these levels or the units could be cooled down to cold re-fueling conditions.
In conclusion, the Units can be brought to the safe shutdown condition even assuming loss of all shutdown related cables in the Auxiliary Building Hallway.
Cable Soreadin Room The normal shutdown cable functions in the Cable Spreading Room are shown in Table 2.
It is noted that these functions are limited to control and instrument cables only and that pump motor power cables, valve motor power
- cables, etc.
are located outside of the Cable Spreading Room.
Loss of function of these instrument and control cables would require plant staff to leave the control room and maintain the units in the hot safe conditions from various 19
13 of 19 locations throughout the plan". Since control cable damage in the spreading room could affect the power operation of important equipment such as cooling water pumps,
- heaters, etc., it would also be necessary to isolate the important shutdown circuits from the remote control locations.
Positioning of control room personnel for performing circuit isolation and control of important equipment is shown in Table 3.
Following completion of the switching and operator positioning at the remote locations, important/requixed hot shutdown functions can be performed.
In the event that remote pressurizer level and steam generator level instrumentation is lost it may be necessary to start the safety injection pumps from the 4160V switchgear.
With the safety injection pumps aligned to the reactor coolant
- system, incxeased auxiliary feedwater/steam dump generator cooling would allow the pressurizer to fill. With the reactor coolant system in the solid condition (at or near the shutoff head pressux'e of the safety injection pumps) and steam generatox heat removal established, the units can be maintained in this condition for a prolonged oeriod of time. It is also noted that the filling of the pressurizer and make-up for xeactor coolant system shrinkage also provides required shutdown boration since the safety injection pumps flow through the boron injection tank.
Cooldown to the residual heat removal mode would not be conducted until (as a
minimum);
(1) at least one level and pressure channel of remote instrumentation for the pressurizer and steam generators is repaired or duplicated, (2) normal reactor coolant system volume control via the charging and chemical volume and control system is re-established and (3) residual heat removal pump control circuits should be repaired or bypassed at the 4160V switchgeax.
Reactox coolant system cooldown should proceed using currently established plant procedures with the assistance of Tables 2 and 3.
As procedures call for positioning of inoperable valves, personnel can be dispatched to make valve adjustments manually.
Once on residual heat removal cooling, reactor coolant 20
]4 of ]9 system conditions can be maintained or cooldown to refueling conditions can be conducted.
Zn conclusion, sufficient capability exists to bring the units to safe shutdown conditions in the event that Cable Spreading Room 'cables are damaged.
21
15 of 19 TABLE 1 SHUTDOWi RELATED CABLE FUNCTIONS IN THE AUXILIARYHALLWAY AT THE 18'LEVATION (p)
{p)
(p)
(P)
(P)
(C)
Charging pumps 3A, 3B
& 3C (C)
Component cooling water pump 4C (P)
Boric acid transfer pumps 3A, 3B, 4A & 4B (P)
Auxiliary feedwater pump steam supply valve's (MOV-3-1405, MOV-4-1405)
Boric acid injection stop valves (MOV-3-350, MOV-4-350)
Volume control tank isolation valve (1!OV-3-115C, MOV-4-115C)
Auxiliary/Radwaste building exhaust and supply fans 3A, 3B, 3A-V10, 3B-Vll (C)
BHR pumps 3A, 3B, 4A & 4B (P)
Accumulator s top valves (i~!OV-3-865C, MOV-4-865A)
(P)
Cold leg safety injection valves (MOV-3-843A, MOV-3-843B, MOV-4-843A)
Feedwater isolation valves (MOV-3-1408, MOV-4-1408)
Residual heat exchanger cooling water isolation valves (MOV-3-749A, MOV-3-749B, MOV-4-749A)
(P)
Refueling water s top valve (MOV-3-862A)
(P)
RHR to RCS isolation valves (MOV-3-744A, MOV-4-744A)
(P)
RCS to RHR isolation valves (MOV-3-751, MOV-4-751)
(I)
Component cooling water flow 3A, 3B, 4A & 4B
~.
(I)
Steam generator level 4A & 4B (P)
Boric acid storage tank heaters A,
B
& C (P)
Boric acid heat tracing transformers A & B (P)
Unit 3
& 4 emergency lighting panels (C)
= Presence of control cable function (P)
= Presence of power cable function
{I) = Presence of instrument cable function 22
16 of 19 TABLE 2 SHUTDOWN RELATED CABLE FUNCTIONS IN THE CABLE SPREADING ROOi~f (c)
(c)
(c)
{c)
(c)
(c)
(c)
(c)
(C)
(c)
{c)
(c)
(c)-
(c)
(c)
(c)
(c)
(C)
(c)
(C)
(c)
(I)
(I)
Charging pumps 3A, 3B, 3C, 4A, 4B
& 4C Component cooling water pumps 3A, 3B, 3C, 4A, 4B
& 4C Intake cooling water pumps 3A, 3B, 3C, 4A, 4B
& 4C Boric acid transfer pumps 3A, 3B, 4A & 4B Boric acid injection stop valves (HOV-3-350, HOV-4-350)
Volume control tank isolation valves (HOV-3-115C, HOV-4-115C)
Auxiliary/Radwaste buildings exhaust and supply fans 3A, 3B 3A-Vlo & 3B-Vll Pressurizer heater control groups 3A & 4A and backup groups 3A, 3B, 4A & 4B RHR Pumps 3A, 3B, 4A & 4B Accumulator stop valves (MOV-3-865A, HOV-3-865B, MOV-3-865C, MOV-4-865A, HOV-4-865B
& HOV-4-865C)
Cold leg SI (HOV-3-843A, MOV-3-843B, iIOV-4-843A & HOV-4-843B)
Main steam isolation valves (POV-3-2604, POV-3-2605, POV-3-2606, POV-4-2604, POV-4-2605
& POV-4-2606)
Feedwater isolation valves (MOV-3-1407, HOV-3-1408, HOV-3-1409, MOV-4-1407, HOV-4-1408
& HOV-4-1409)
RHR heat exchanger cooling water isolation valves (HOV-3-749A, HOV-3-749B, 'i'IOV-4-749A & MOV-4-749B)
Refueling water storage tank stop valves (HOV-3-862A, MOV-3-862B, i~fOV-4-862A, MOV-4-862B)
RHR to RCS isolation valves (HOV-3-744A, i~fOV-3-744B, ifOV-4-744A
& HOV-4-744B)
RCS to RHR isolation (MOV-3-750, MOV-4-750, i~fOV-3-751, HOV-4-75 1)
Auxiliary feedwater pump steam supply valves (HOV-3-1403,
'LLIOV-3-1404 I MOV-3-1405, ifOV-4-1403, HOV-4-14041 ifOV-4-1405)
Unit 3
& 4 auxiliary feedwater auto start and backup circuits Auxiliary feedwater-pump steam pressure controllers (SV-3706, S V-3707)
Auxiliary feedwater control and backup valves, 3A, 3B, 3C, 4A, 4B
& 4C RCS pressure (PX-3-403, PI-4-403)'CS temperature (hot let 3A, 3B, 3C, 4A, 4B, 4C; cold leg 3A, 3B, 3C, 4A, 4B, 4C) 23
of 19 TABLE 2 (cont)
(I)
(~)
(c)
(G)
Steam generator pressure and level 3A, 3B, 3C, 4A,, 4B, 4C Unit 3 6 4 pressurizer level 480 V load centers 3A, 3B, 3C, 3D, 4A, 4B, 4C 6 4C feeder breakers Diesel generator. breakers (generators f33
& 4; buses 3A, 4A, 3B 6 4B)
(C)
= Presence of control cable function (P)
~ Presence of power cable function (l') = Presence of instrument cable function 24
TABLE 3 EXAMPLE OF PLANT STAFF ACTIONS NECESSARY. TO CONTROL AND MAINTAIN HOT CONDITIONS FRON RE~iIOTE PLANT LOCATIONS (1)
Nuclear Control Center 0 erator (1 of 3)
Place Isolation Switches In Remote At The Followin Locations:
4160 Volt 4A Switch ear Room Location 480V Load Center 4A Bus tie to switchgear 4B Component cooling water pump 4A 480V Load Center 4C Intake cooling water pump 4A
~ Emergency diesel generator A
Isolation switch on "A" sequencer panel 4160 Volt 4B Switch ear Room Location 480V Load Center 4B Component cooling pump 4B 480V Load Center 4D Intake cooling water pump 4B Intake cooling water pump 4C Emergency diesel generator B
Bus tie to 4A switchgear Isolation switch on "B" sequencer panel Load Center Room Location, Unit 4 (south wall)
Charging pumps 4A, 4B, 4C NCC 4A - North End Location Normal containment cooler 4C Place control switch on fast Cable S readin Room Location Isolation switch, Isolation switch, Isolation switch, Isolation switch, generator s~3 lockout relay Unit 3 startup transformer lockout generator
/f4 lockout relay Unit 4 startup transformer lockout relay relay DC Switch ear Room Location Isolation switch, DC emergency bearing oil pump Unit on 125V DC panel A
Isolation switch, DC emergency bearing oil pump Unit on 125V DC panel B
3 j 4,
<<Note that while it is perfered to isolate these lockout relays it is not absolutely essential since the equipment required for shutdown can be powered from the emergency diesel generators.
25
19 of 19 TABLE 3 (cont)
(2)
Nuclear Natch En ineer Place Isolation Switches In Remote At The Followin Locations:
4160 Volt 3A Switch ear Room Location 480V Load Center 3A Bus tie to switchgear 3B Component cooling water pump 3A 480V Load Center 4C Component cooling pump 3C Intake cooling water pump 3A Emergency diesel generator A
Isolation switch on "A" sequencer panel 4160 Volt 3B Switch ear Room Location 480V Load Center 3B Component cooling pump 3B 480V Load Cen ter 3D Intake cooling water pump 3B Intake cooling water pump 3C Emergency diesel generator B
Bus tie to 3A switchgear Isolation switch'n "B" sequencer panel Load Center Room Location Unit 3 (south wall)
Charging pumps 3A, 3B, 3C
~ULCC 4A North End Location Normal containment cooler 3A Place control switch on fast Emer enc Diesel Generator Buildin Diesel generator A control panel Diesel generator B control panel (3)
Nuclear Turbine 0 erator Proceed to the Unit 3 auxiliary feedwater control station and maintain Unit 3 steam generator levels (4)
Nuclear Control Center 0 erator (1 of 3)
Proceed to the Unit 4 auxiliary feedwater control station and maintain Unit 4 steam generator levels (5)
Nuclear Control Center 0 erator (1 of 3)
Proceed to Unit 4 charging pump room and maintain pressurizer level as required and perform boration of RCS as directed (6)
Nuclear 0 erator Proceed to Unit 3 charging pump room and maintain pressurizer level as required and perform boration of RCS as directed 26
ATTACHMENT 2 FPL
SUMMARY
OF THE SANDIA TESTS WITH REGARD TO FLAMEMSTIC 71A 27
m
~
~
FLORIDA POWER AND LIGHT CO~PAVE'S SHMARY OF THE SANDIA TESTS WITH REGARD TO FLAHEMASTIC 71A The full scale two-tray coating fire tests conducted by Sandia were established to determine if flame propagatio'n would occur between trays as a result of a Eire in the bottom tray.
The Sandia test method is probably the most severe test method used to date because of the conservative test conditions, e.
g.
1.).
Open ladder cable trays, 2.)
Very low cable fillpercent because of figure eighting of cables in order to obtain heat and air venting openings, 3.)
Used the worst of the unqualified cables, 4.)
Used two 70,000 btu/hr burners and 5.)
Both burners aligned to directly impact the coated cables a distance of 4.75 inches away.
In the Flamemastic 71A, non-IEEE 383 cable test, it took 10 minutes to ignite the cables in the bottom tray.
Following this, the cables in the bottom tray burned for 26 miniites before self-extinguishing and achieved a maximum temper-ature of only 1260 F. The cables in the top tray did not short, did not burn, were undamaged and achieved a maximum temperature of only 167 F.
In conclu-0
- sion, the same coating system (Flamemastic 71A), with the same application technique, on cables similar to those at Turkey Point and St. Lucis passed the Sandia test.
Please note that a larger exposure fire than that used in the ~st would not have caused propagation of flames to the upper tray, but would have only ignited the lower tray sooner.
Further and contrary to the Staff's position, Flamemastic 71A also effectively seals in the combustible inventory associated with the cable insulation.
The very severe Sandia fire test of the Flamemastic 71A coated non-IEEE 383 cables discussed above yielded a weight loss of only 7-1/2 pounds.
This loss is attributable to burned cable insulation and pyrolized organic binder in the Flamemastic 71A and in total has a heat release equivalent of approximately two changes of street clothing.
This small amount of energy spread over the fire areas in question will have an insignificant affect on temperatures.
28