ML061770163

Browns Ferry - Units 1, 2 and 3 - Technical Specifications (TS) Changes TS-431 & TS-418 Extended Power Uprate (EPU) - Revised Response to NRC Round 2 Requests for Additional Information SPLB-A.1, SPLB-A.2, & SPLB-A.3

ML061770163
Person / Time
Site:	Browns Ferry
Issue date:	06/23/2006
From:	Crouch W D Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
TAC MC3743, TAC MC3744, TAC MC3812, TVA-BFN-TS-418, TVA-BFN-TS-431
Download: ML061770163 (15)

Text

June 23, 2006

TVA-BFN-TS-431

TVA-BFN-TS-418

10 CFR 50.90

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk

Mail Stop OWFN, P1-35

Washington, D. C. 20555-0001

Gentlemen:

In the Matter of ) Docket Nos. 50-259 Tennessee Valley Authority ) 50-260

) 50-296

BROWNS FERRY NUCLEAR PLANT (BFN) - UNITS 1, 2, AND 3 -

TECHNICAL SPECIFICATIONS (TS) CHANGES TS-431 AND TS-418 -

EXTENDED POWER UPRATE (EPU) - REVISED RESPONSE TO NRC ROUND 2

REQUESTS FOR ADDITIONAL INFORMATION SPLB-A.1, SPLB-A.2, AND

SPLB-A.3 - (TAC NOS. MC3812, MC3743, AND MC3744)

By letter dated April 13, 2006 (ADAMS Accession No.

ML061040217), TVA submitted revised responses to NRC Round 2

requests for additional information (RAIs) regarding TVA's

applications for extended power uprate of BFN Units 1, 2 and

3. As stated in that letter, Item 3.1.2(5) of the

supplemental reply to RAIs SPLB-A.1, SPLB-A.2, and SPLB-A.3

was not complete at that time. The enclosure to this letter

provides the completed response to the subject RAIs. The

response provided is the same for all three BFN units.

U.S. Nuclear Regulatory Commission Page 2 June 23, 2006

If you have any questions regarding this letter, please

contact me at (256)729-2636.

I declare under penalty of perjury that the foregoing is true

and correct. Executed on this 23 rd day of June, 2006.

Sincerely,

Original signed by:

William D. Crouch

Manager of Licensing

and Industry Affairs

Enclosure:

Supplement to Response to NRC Round 2 Requests for Additional Information - SPLB-A.1, SPLB-A.2, and SPLB-A.3 cc: (see page 3)

U.S. Nuclear Regulatory Commission Page 3 June 23, 2006

Enclosure

cc (w/Enclosure):

State Health Officer

Alabama Dept. of Public Health

RSA Tower - Administration

Suite 1552

P.O. Box 303017

Montgomery, AL 36130-3017

U.S. Nuclear Regulatory Commission

Region II

Sam Nunn Atlanta Federal Center

61 Forsyth Street, SW, Suite 23T85

Atlanta, Georgia 30303-3415

Malcolm T. Widmann, Branch Chief

U.S. Nuclear Regulatory Commission

Region II

Sam Nunn Atlanta Federal Center

61 Forsyth Street, SW, Suite 23T85

Atlanta, Georgia 30303-8931

NRC Senior Resident Inspector

Browns Ferry Nuclear Plant

10833 Shaw Road

Athens, Alabama 35611-6970

NRC Unit 1 Restart Senior Resident Inspector

Browns Ferry Nuclear Plant

10833 Shaw Road

Athens, Alabama 35611-6970

Margaret Chernoff, Project Manager

U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North

11555 Rockville Pike

Rockville, Maryland 20852-2739 Ms. Eva A. Brown, Project Manager U.S. Nuclear Regulatory Commission (MS 08G9)

One White Flint, North 11555 Rockville Pike Rockville, Maryland 20852-2739

U.S. Nuclear Regulatory Commission Page 4 June 23, 2006

DTL:LTG:BAB

cc (w/o Enclosure):

B. M. Aukland, POB 2C-BFN

M. Bajestani, NAB 1A-BFN

A. S. Bhatnagar, LP 6A-C

J. C. Fornicola, LP 6A-C

R. G. Jones, POB 2C-BFN

G. V. Little, NAB 1A-C

R. F. Marks, Jr., PAB 1C-BFN

G. W. Morris, LP 4G-C

B. J. O'Grady, PAB 1E-BFN

K. W. Singer, LP 6A-C

E. J. Vigluicci, ET 11A-K

NSRB Support, LP 5M-C

EDMS WT CA-K, w.

s:lic/submit/TechSpec/TS 418 and 431 - Revised RAIs.doc

E-1 ENCLOSURE TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNITS 1, 2, AND 3 TECHNICAL SPECIFICATIONS (TS) CHANGE NOS. TS-418 AND TS-431 -

SUPPLEMENT TO RESPONSE TO NRC ROUND 2 REQUESTS FOR ADDITIONAL INFORMATION - SPLB-A.1, SPLB-A.2, AND SPLB-A.3 By letter dated April 13, 2006 (ADAMS Accession No. ML061040217),

TVA submitted revised responses to NRC Round 2 requests for

additional information (RAIs) regarding TVA's applications for

extended power uprate of BFN Units 1, 2 and 3. As stated in that

letter, Item 3.1.2(5) of the supplemental reply to RAIs SPLB-A.1, SPLB-A.2, and SPLB-A.3 was not complete at that time. This

enclosure provides the completed response to the subject RAIs.

The response is the same for all three BFN units. The entire

response to these RAIs is provided below for completeness;

however, only the response to Item 3.1.2(5) has been revised and

is marked by revision bar in the page margin.

NRC Request SPLB-A.1 Section 10.5.5 of the Updated Final Safety Analysis Report (UFSAR), Revision 17 dated August 30, 1999, revised the

discussion from the UFSAR that was previously provided regarding

the maximum SFP heat load for batch and full core offloads. In

order to facilitate NRC review of the capability of the SFPCCS to

perform its function for EPU conditions, provide a discussion on

the safety-related systems required to maintain fuel pool cooling

within design bases temperature limits.

NRC Request SPLB-A.2 For EPU conditions, explain how the SFP water temperature will be maintained below 150 degrees Fahrenheit (F) for the worst-case

normal (batch) and full core offload scenarios assuming a loss of

offsite power and (for the batch offload only) a concurrent

single active failure considering all possible initial

configurations that can exist. Include a description of the

maximum decay heat load that will exist in the SFP for each case, how these heat loads were determined, such that they represent

the worst-case conditions, and what the cooling capacity is for

the systems that are credited, including how this determination

was made. Also: a. Describe any operator actions that are required, how long it will take to complete these actions, and how this

determination was made; and E-2 b. Describe the maximum core decay heat load that will exist at the onset of fuel movement, how this determination was made, how this heat load will be accommodated while also

satisfying the SFP cooling requirements over the duration of

the respective fuel offload scenarios, and including the

situation where the SFP is isolated from the reactor vessel

cavity. NRC Request SPLB-A.3 Discuss how adequate SFP makeup capability is assured for EPU conditions in the unlikely event of a complete loss of SFP

cooling capability, including how the maximum possible SFP

boil-off rate compares with the assured makeup capability that

exists, operator actions that must be taken, how long it will

take to complete these actions and how this determination was

made, and boron dilution considerations.

TVA's Supplemental Reply to SPLB-A.1, 2, and 3 TVA has previously provided information regarding the spent fuel pool cooling system at BFN and the effects of EPU in PUSAR

Section 6.3 and in the December 19, 2005, reply to questions

SPLB-A.1, SPLB-A.2, and SPLB-A.3. Additional discussion was

provided in the April 13, 2006 letter to clarify and provide

supplemental information on the BFN spent fuel pool cooling

system and was presented in the format (including numbering) of to Matrix 5 of RS-001, "Review Standard for Extended

Power Uprates," Revision 0, December 2003.

At the time of the April 13, 2006 letter, information was not available to complete the response to Item 3.1.2(5). The

following discussion provides the remaining information.

Revisions to the previous response dated April 13, 2006 are

marked with revision bars.

1. BACKGROUND The BFN fuel pool cooling and cleanup systemsfor Units 1, 2, and 3 are described in UFSAR Section 10.5. The systems cool

the fuel storage pools by transferring the spent fuel decay

heat through heat exchangers to the reactor building closed

cooling water (RBCCW) systems. The system for each fuel pool

consists of two circulating pumps connected in parallel, two

heat exchangers, one filter demineralizer subsystem, two

skimmer surge tanks, and the required piping, valves, and

instrumentation. Four filter demineralizers are provided

including one spare filter demineralizer shared between the

three units. The pumps circulate the pool water in a closed

loop, taking suction from the surge tanks, circulating the

water through the heat exchangers and filter demineralizer, and discharging it through diffusers at the bottom of the E-3 fuel pool and reactor well (as required during refueling operations). The water flows from the pool surface through

skimmer weirs and scuppers (wave suppressers) to the surge

tanks.

The heat exchangers in the residual heat re m oval (RHR) system can be used in conjunction with the fuel pool cooling and

cleanup system to supplement pool cooling (supplemental fuel

pool cooling). Normal makeup water for the f uel pool cooling system is transferred from the condensate storage tank t o the skimmer surge tanks. A seismic Class I qualified source of

makeup water is provided through the crosstie between the RHR

system and fuel pool cooling system. If necessary, the

intertie between the RHR service water (RHRSW) system and the

RHR system can be utilized to admit raw water as makeup.

Also, a standpipe and hose connection is provided on each of

the two emergency equipment cooling water (EECW) system

headers which provide two additional fuel pool water makeup

sources.

Additionally, the auxiliary decay heat removal (ADHR) system

provides another means to remove decay heat and residual heat

from the spent fuel pool and reactor cavity of BFN Units 2

and 3 and is described in UFSAR Section 10.22. As part of

restart activities for BFN Unit 1, the ADHR system will be

extended to include the spent fuel pool and reactor cavity of

BFN Unit 1. During operation of this system, it is aligned

to only one unit at a time. The ADHR system consists of two

cooling water loops. The primary cooling loop circulates

spent fuel pool water entirely inside the Reactor Building

and rejects heat to a secondary loop by means of a heat

exchanger. The secondary loop transfers heat to the

atmosphere outside the Reactor Building by means of

evaporative cooling towers.

Spent fuel pool cooling, including supplemental fuel pool

cooling and ADHR, are non-safety systems. To ensure adequate

makeup under all normal and off normal conditions, the

RHR/RHRSW connection provides a permanently installed seismic

Class I qualified makeup water source for the spent fuel

pool. This ensures that irradiated fuel is maintained

submerged in water and that reestablishment of normal fuel

pool water level is possible under all anticipated

conditions. Two additional sources of spent fuel pool water

makeup are provided via a standpipe and hose connection on

each of the two EECW headers. Each hose is capable of

supplying makeup water in sufficient quantity to maintain

fuel pool water level under conditions of no fuel pool

cooling.

E-4 2. ACCEPTANCE CRITERIA The current design and operational basis for BFN spent fuel pool cooling system is as follows: Administrative controls are used to ensure that the fuel pool heat load does not exceed available cooling capacity. The capacity of the spent fuel pool cooling and the ADHR systems, considering seasonal cooling water temperatures

and current heat exchanger conditions, are utilized to maintain the fuel pool temperature at or below 125 F during normal refueling outages (average spent fuel batch

discharged from the equilibrium fuel cycle). The RHR system can be operated in parallel with the spent fuel pool cooling system to maintain the fuel pool

temperature less than the Technical Requirements Manual (TRM) limit of 150F if a full core off load is performed.

Plant instructions require that actions be taken well

before exceeding this limit. The fuel pool temperature is normally maintained between 72 F and 125 F. To ensure adequate makeup under all normal and off normal conditions (i.e. fuel pool water boil off), the RHR/RHRSW

crosstie provides a permanently installed seismic Class I

qualified makeup water source for the spent fuel pool. Two additional sources of spent fuel pool water makeup are provided via a standpipe and hose connection on each of

the two EECW headers. Each hose is capable of supplying

makeup water in sufficient quantity to maintain fuel pool

water level under conditions of no fuel pool cooling.

The design basis for the fuel pool cooling systems remains

the same for the current and EPU conditions.

3. REVIEW PROCEDURES

3.1 Adequate

SFP Cooling Capacity To demonstrate adequate SFP cooling capacity, BFN performs both bounding and cycle-specific calculations. The bounding

calculations have been reperformed for EPU conditions as

described below in Section 3.1.1 to ensure that the

acceptance criteria will continue to be met. Additionally, as described in Section 3.1.2, cycle-specific calculations

are performed to assess cooling system capability to ensure

that fuel pool heat load does not exceed available cooling

capacity. These calculations demonstrate that the

acceptance criteria described in Section 2 will continue to

be met under EPU conditions.

E-5 As a result of EPU, the normal spent fuel pool heat load

will be higher than the pre-EPU heat load. EPU will result

in higher decay heat in the discharged bundles to the spent

fuel pool as well as an increase in the number of discharged

fuel bundles at the end of each cycle. The heat removal

capability of the spent fuel pool cooling system, the ADHR

system, or the supplemental fuel pool cooling mode of the

RHR system are not affected by EPU. The evaluations for

spent fuel pool cooling, as discussed below, include the

effects from EPU operation and provide the results

indicating that the design basis for the spent fuel pool

will be maintained.

3.1.1 Bounding

Calculation Consistent with the BFN design basis, two cases were analyzed: 1. Partial core offload with operation of the

spent fuel pool cooling system and ADHR system, and

2. Full core offload with operation of the spent fuel pool

cooling system and RHR supplemental fuel pool cooling

mode. In each case the initial fuel pool temperature was

assumed to be 100°F.

1. Partial Core Offload The capacity of the fuel pool cooling system and the ADHR system to maintain the fuel pool temperature at or

below 125°F during partial core offloads was evaluated

for EPU conditions.

The maximum decay heat loadings for the spent fuel pool

were calculated using the ANSI/ANS 5.1-1979 Standard

with two-sigma uncertainty. The heat load in the spent

fuel pool is the sum of previous fuel offloads and the

recent batch decay heats at the time of transfer. In

this analysis, the offload consists of a batch of 332

fuel bundles offloaded to an almost full spent fuel

pool. This batch size was chosen for analytical

purposes; the actual batch size may vary.

The spent fuel pool was assumed to be previously loaded

with 2375 bundles allowing a reserve space for a full

core offload (764 cells). The 2375 bundles were

assumed to have been offloaded in eight batches, discharged at 24 month intervals. For this case, core

offload begins 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after reactor shutdown. Fuel

transfer time was estimated based on a transfer rate of

14 bundles per hour to the fuel pool. These decay heat

and offload time estimates establish the limiting case

maximum heat loads.

E-6 Cooling of the fuel pool conservatively assumes that

only one heat exchanger/pump combination is available

for each system. The heat exchanger effectiveness is

based upon original design specifications including

standard value fouling factors and tube plugging

criteria. The evaluation only considers the mass of

water in the fuel pool and assumes no circulation of

water between the fuel pool and the cavity for the

period of time that fuel pool gates are open while the

fuel is being transferred to the pool.

The results of this evaluation show that the peak spent

fuel pool temperature remains less than 125°F under EPU

conditions.

Table 1 Partial Core Offload Evaluation Results for One Train Each of Spent Fuel Pool Cooling System and ADHR 1 Conditions/Parameters Value Peak spent fuel pool temperature (°F) 99.1 Time to peak spent fuel pool temperature (hours) 80 Time to boil from loss of all cooling at peak temperature (hours) 14 Boil off rate (gpm) 48 1 Assumes core offload begins 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after reactor shutdown to allow for cooldown, vessel head removal, refueling cavity filling, and other refueling

preparations.

PUSAR Table 6-3 contains an additional case where a partial core offload was evaluated for one train each

of the spent fuel pool cooling system and RHR

supplemental fuel pool cooling mode. In that

evaluation, the calculated peak spent fuel pool

temperature of 124.9°F was less than 125°F.

E-7 Table 2 Partial Core Offload Evaluation Results for One Train Each of Spent Fuel Pool Cooling System and RHR Supplemental Fuel Pool Cooling Mode 1 Conditions/Parameters Value Peak spent fuel pool temperature (°F) 124.9 Time to peak spent fuel pool temperature (hours) 130 Time to boil from loss of all cooling at peak temperature (hours) 13 Boil off rate (gpm) 42 1 Assumes core offload begins 95 hours0.0011 days <br />0.0264 hours <br />1.570767e-4 weeks <br />3.61475e-5 months <br /> after reactor shutdown and includes 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of invessel stay time because the RHR supplemental fuel pool cooling mode has

less heat removal capacity than the ADHR.

2. Full Core Offload The capacity of the spent fuel pool cooling system and the RHR supplemental fuel pool cooling mode to maintain the fuel pool temperature at or below 150 F during a full core off load is evaluated for EPU conditions.

The maximum decay heat loadings for the spent fuel pool

were calculated using the ANSI/ANS 5.1-1979 Standard with

two-sigma uncertainty. The heat load in the spent fuel

pool is the sum of previous fuel offloads and the recent

full core decay heats at the time of transfer. The pool

is assumed to be previously loaded with 2707 bundles.

The prior offload batches were assumed to be the same as

the partial core offload case above with an additional

batch of 332 fuel assemblies having been discharged from

the reactor core, all of which has been cooled for an

additional 24 months. (The partial offload batch size was chosen for analytical purposes; the actual may vary.)

The initiation of fuel offloading was a minimum of 50

hours after plant shutdown based upon shutdown cooling

requirements, head removal time and refueling

preparation. Actual times were determined based on the

calculated heat removal capacity of the cooling mode.

For this case, core offload begins 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> after

reactor shutdown and includes 115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> of invessel stay

time because the RHR supplemental fuel pool cooling mode E-8 has less heat removal capacity than the ADHR system.

Fuel transfer time was estimated based on a transfer rate

of 14 bundles per hour to the fuel pool. These decay

heat and offload time estimates establish the limiting

case maximum heat loads.

Cooling of the fuel pool conservatively assumes that only

one heat exchanger/pump combination is available for each

system. The heat exchanger effectiveness is based upon

original design specifications including standard value

fouling factors and tube plugging criteria. The

evaluation only considers the mass of water in the fuel

pool and assumes no circulation of water between the fuel

pool and the cavity for the period of time that fuel pool

gates are open while the fuel is being transferred to the

pool.

The results of this evaluation show that the peak spent

fuel pool temperature remains less than 150°F under EPU

conditions.

Table 3 Full Core Offload Evaluation Results for One Train Each of Spent Fuel Pool Cooling System and RHR Supplemental Fuel Pool Cooling Mode 1 Conditions/Parameters ValuePeak spent fuel pool temperature (°F) 149.8Time to peak spent fuel pool temperature (hours) 229 Time to boil from loss of all cooling at peak temperature (hours) 4 Boil off rate (gpm) 80 1 Assumes core offload begins 165 hours0.00191 days <br />0.0458 hours <br />2.728175e-4 weeks <br />6.27825e-5 months <br /> after reactor shutdown and includes 115 hours0.00133 days <br />0.0319 hours <br />1.901455e-4 weeks <br />4.37575e-5 months <br /> of invessel stay time because the RHR supplemental fuel pool cooling mode has

less heat removal capacity than the ADHR.

PUSAR Table 6-3 contains an additional case where a full core offload was evaluated for one train each of

the spent fuel pool cooling system and ADHR system. In

that evaluation, the calculated peak spent fuel pool

temperature of 121.5°F was also less than 150°F.

E-9 Table 4 Full Core Offload Evaluation Results for One Train Each of Spent Fuel Pool Cooling System and ADHR 1 Conditions/Parameters ValuePeak spent fuel pool temperature (°F) 121.5Time to peak spent fuel pool temperature (hours) 109 Time to boil from loss of all cooling at peak temperature (hours) 5 Boil off rate (gpm) 104 1 Assumes core offload begins 50 hours5.787037e-4 days <br />0.0139 hours <br />8.267196e-5 weeks <br />1.9025e-5 months <br /> after reactor shutdown to allow for cooldown, vessel head removal, refueling cavity filling, and other refueling

preparations.

3.1.2 Cycle-Specific Calculation Unloading the reactor core and the associated increase in fuel pool heat load is a controlled evolution.

Administrative controls are used to ensure that the fuel

pool heat load does not exceed available cooling capacity, such that the fuel pool gates are not closed until the

decay heat load is less than or equal to the fuel pool

cooling heat exchanger capacity. Performance of the fuel

pool cooling systems is predicted prior to each refueling

outage as part of the Outage Risk Assessment Review (ORAM)

process.

In addition to the following discussion, BFN is taking

additional actions to further augment procedures

pertaining to the cycle specific administrative controls.

Procedure changes will be generated (1) to define and

control the generation of cycle-specific fuel pool heat

load calculations, and (2) to control the installation of

the fuel pool gates based on the calculated fuel pool heat

load.

Cycle-specific analysis conditions:

(1) Predicted decay heat for both the spent fuel pool and reactor core are determined by utilizing a TVA code (DHEAT) that complies with the methods of ANSI/ANS

5.1. The history of previous fuel discharges is used

as input into the decay heat load determination for

the spent fuel pool. The decay heat results are best-E-10 estimate values and are provided for a range of decay times that may be needed for the spent fuel pool

evaluations.

(2) Cooling system heat removal is calculated utilizing a spreadsheet based on heat balances of the affected

systems. Fuel pool cooling capacity of the systems is

based upon inlet cooling temperatures, system flow

rates, trains in service, and heat exchanger

performance values.

(3) As described in (2) above, heat removal capabilities are determined for each of the BFN cooling trains, including the normal spent fuel pool cooling system, the ADHR system, and the supplemental fuel pool

cooling mode of RHR.

(4) The limiting parameter for heat load and heat removal capability is the insertion of the fuel pool gates

following core offload. When the fuel pool gates are

removed and spent fuel movement begins, additional

cooling is provided by the shutdown cooling system

that provides decay heat removal directly to the

reactor vessel. Evaluations of the spent fuel pool

temperature following discharge of the partial core

offload are performed based on cooling system

configurations to ensure that the spent fuel pool

temperature can be maintained without the additional

heat removal capacity of the shutdown cooling system.

(5) Calculations of spent fuel pool heat load and heat removal capability as described herein are premised

with maintaining the spent fuel pool temperature

within the Technical Requirements Manual (TRM) 3.9.2

limit of <

150°F based on limited equipment availability. Additionally, plant instructions

require that actions be taken well before exceeding

this value.

The spent fuel pool structure will accommodate boiling

conditions in the case of a complete loss of spent

fuel pool cooling. In this case, adequate make-up

supply is available as described in Section 3.2 below (6) Administrative controls are provided as part of ORAM to ensure that appropriate controls are provided for

shutdown safety. These controls ensure proper

assessment of key shutdown areas (i.e., reactivity

control, shutdown cooling, AC power, fuel pool

cooling, etc.). Spent fuel pool cooling assessments E-11 are performed prior to the outage and updated during the outage to ensure appropriate controls are

maintained for the safe operation of spent fuel pool

cooling. 3.2 Adequate Make-Up Supply The evaluations described in Sections 3.1.1.1 and 3.1.1.2 above are used to determine the time to boil for make-up

capability. These evaluations assume only one train of each

cooling system is in operation to determine the peak spent

fuel pool temperature. At the time of peak spent fuel pool

temperature, it is assumed that all spent fuel pool cooling

is lost. Based on decay heat, the time to reach boiling

conditions is then calculated. The results are provided in

Tables 1 through 4 above.

The minimum time to reach boiling is four hours based on the

case presented in Table 3. This case involves a full core

offload and assumed loss of all cooling at the peak spent

fuel pool temperature of 149.8°F. The associated boil off

rate is 80 gpm.

The maximum boil off rate is 104 gpm based on the case

presented in Table 4. This case involves a full core

offload and assumed loss of all cooling at the peak spent

fuel pool temperature of 121.5°F. The associated time to

reach boiling is five hours.

For BFN the RHR/RHRSW crosstie provides a permanently installed seismic Class I qualified makeup water source for

the spent fuel pool. This supply can be aligned within the

minimum four hours calculated above and can supply greater

than 150 gpm to the spent fuel pool.

Two additional sources of spent fuel pool water makeup are

provided via a standpipe and hose connection on each of the

two EECW headers. Each hose is capable of supplying makeup

water at 150 gpm to the spent fuel pool within the minimum

four hours calculated above.