ML061040217

Technical Specifications (TS) Change Nos. TS-418 and TS-431- Extended Power Uprate (EPU) Operation - Revised Responses to NRC Round 2 Requests for Additional Information

ML061040217
Person / Time
Site:	Browns Ferry
Issue date:	04/13/2006
From:	Crouch W 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: ML061040217 (22)

Text

April 13, 2006

TVA-BFN-TS-418

TVA-BFN-TS-431

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) CHANGE NOS. TS-418 AND TS-431 -

EXTENDED POWER UPRATE (EPU) OPERATION - REVISED RESPONSES TO

NRC ROUND 2 REQUESTS FOR ADDITIONAL INFORMATION - (TAC NOS.

MC3812, MC3743, AND MC3744)

By letters dated December 19, 2005 (ADAMS Accession Nos.

ML053560194 and ML053560186) TVA submitted responses to NRC

Round 2 requests for additional information (RAIs) regarding

TVA's applications for extended power uprate of BFN Unit 1 and

BFN Units 2 and 3, respectively. As a result of discussions

with the NRC staff, TVA is revising its responses to five of

the RAIs. The responses to the subject RAIs are the same for

all three BFN units.

to this letter provides revised responses to RAIs

IPSB-B.1 and IPSB-B.8 regarding external radiation doses due

to direct radiation and skyshine. Enclosure 2 to this letter

provides revised responses to RAIs SPLB-A.1, SPLB-A.2 and

SPLB-A.3 regarding fuel pool cooling. Each revised RAI

response in Enclosure 1 supersedes the response previously

provided to the NRC staff. However, the responses in supplement the previous responses to the

respective RAIs.

U.S. Nuclear Regulatory Commission Page 2 April 13, 2006 Item 3.1.2(5) of the Supplemental Reply to RAIs SPLB-A.1, 2, and 3 in Enclosure 2 is not complete. As discussed with the

NRC staff on April 13, 2006, TVA will address this item in a

supplemental reply.

identifies a regulatory commitment made in to modify the administrative controls regarding

spent fuel pool cooling operations.

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 13 th day of April, 2006.

Sincerely,

Original signed by:

William D. Crouch

Manager of Licensing

and Industry Affairs

Enclosures

1. Revised Responses to RAIs IPSB-B.1 and IPSB-B.8

2. Supplements to Responses to RAIs SPLB-A.1, SPLB-A.2 and SPLB-A.3 3. Commitment Listing

cc (See page 3.)

U.S. Nuclear Regulatory Commission Page 3 April 13, 2006

Enclosures:

cc (Enclosures):

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

U.S. Nuclear Regulatory Commission Page 4 April 13, 2006 JEM:LTG:BAB Enclosures

cc (Enclosures):

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

E1-1 ENCLOSURE 1 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

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

REVISED RESPONSES TO NRC ROUND 2 REQUESTS FOR ADDITIONAL INFORMATION -

IPSB-B.1 and IPSB-B.8 By letters dated December 19, 2005 (ADAMS Accession Nos.

ML053560194 and ML053560186) TVA submitted responses to NRC

Round 2 requests for additional information (RAIs) regarding

TVA's applications for extended power uprate of BFN Unit 1 and

BFN Units 2 and 3, respectively. As a result of discussions with

the NRC staff, TVA is superseding its prior responses to RAIs

IPSB-B.1 and IPSB-B.8 regarding external radiation doses due to

direct radiation and skyshine. The responses to the subject RAIs

are the same for all three BFN units.

NRC Request IPSB-B.1 Section 8.6, Normal Operations Off-Site Doses, of Enclosure 4 of the [June 25, 2004 and June 28, 2004 submittals for BFN Units 2

and 3 and BFN Unit 1, respectively] states that radiation from

shine (offsite) is not presently a significant exposure pathway

and is not significantly affected by EPU. This conclusion is

based on the experience of earlier 5-percent power uprates for

Units 2 and 3. Also, Section 8.2.2, Offsite Doses at Power

Uprate Conditions, of the Environmental Report states that N-16

activity in the Turbine Building will increase linearly with EPU.

The magnitude of the N-16 source term in the Turbine Buildings is

not a simple linear increase with reactor power. The equilibrium

concentration of N-16 in the Turbine Building systems will be

effected (an inverse exponential function) by the decreased decay

resulting from the increased steam/feed flow between the reactor

and the Turbine Building. Implementation of hydrogen injection

water chemistry also increases N-16 concentrations in reactor

steam independently of reactor power.

Provide the present nominal value for the skyshine external dose

component (assuming all three units operating at current licensed

power levels), the corresponding estimated dose component

following EPU (assuming all three units operating at the

requested power, and design basis steam activity, levels).

Include all parameters (i.e., flow rates, system component

dimensions, etc.) used in calculating these values and specify

the calculational method used. Identify the limiting dose

receptor (i.e., is the dose receptor a member of the public E1-2 located offsite and, therefore, subject to the dose limits of 40 CFR Part 190) or a member of the public working onsite (subject to the dose limits of 20.1301)). Describe any increases

in doses for onsite spaces (i.e., Administrative offices, guard

stations, etc.) continuously or routinely occupied by plant

visitors or staff.

TVA Reply to IPSB-B.1 A number of studies have been conducted at BFN to characterize the direct radiation and building/atmospheric scatter skyshine

radiation fields associated with increased N-16 and C-15

production from hydrogen injection into the feedwater system for

mitigation of intergrannular stress corrosion cracking (IGSCC) of

vessel internals. Radiation levels onsite have been measured

with thermoluminescent dosimeters (TLD), pressurized ionization

chambers (PIC), and hyper-pure germanium detectors.

In 1997, while Units 2 and 3 were operating at original licensed

thermal power (OLTP) (i.e., 3293 MWt), and prior to the injection

of hydrogen in the feedwater system on either unit, GE Nuclear

Energy performed extensive surveys of site radiation levels. A

subsequent report, GE-NE-P7300044-01-01-00, "Browns Ferry Nuclear

Power Station, Potential Dose Consequences Resulting From

Implementation of Hydrogen Water Chemistry," provides dose rate

projections through a number of shield wall thicknesses as a

function of distance from BFN units under OLTP and normal water

chemistry conditions. The projections were based on the output

from the mathematical model provided in "BWR Turbine Equipment

N-16 Radiation Shielding Studies," by D. R. Rogers, General

Electric NEDO-20206, 1973, normalized to a PIC radiation

measurement in line with the operating turbines at the north end

of the electrical switchyard. This location was chosen because

it is unaffected by other sources of radiation such as from

radwaste processing/shipment or the condensate storage tanks.

The report provided projections in-line with the turbines and

normal to them. As the former were slightly more conservative (i.e., provided higher values), they were used for dose rate

projections in occupied areas onsite. The dose rate projection

curves in the report extend to 2000 feet from the turbine center

line. The projection curves were then extrapolated in order to

project site boundary doses. The dose rate at the nearest site

boundary (i.e., 3850 feet) was projected to be 0.04 µR/h per unit

under OLTP and normal water chemistry conditions. This equates

to a total annual dose of approximately 1.1 mrem from all three

units at OLTP.

Components on the turbine deck which contribute to the skyshine

include: the piping to and from the high pressure turbine, the

high pressure turbine, the crossover piping from the moisture

separators to the low pressure turbines, the combined intercept E1-3 valves and the low pressure turbines. According to the GE report (GE-NE-P7300044-01-01-00), the vast majority (~72%) of the

skyshine emanates from N-16 and C-15 in the steam traversing the

crossover piping. Based on this, the change in travel time of

the steam to the midpoint of the crossover piping was calculated

for OLTP, current licensed thermal power (CLTP) (i.e., 3458 MWt)

and EPU (i.e., 3952 MWt) conditions. Steam travel times were

calculated based on the steam flow rates for each of the power

levels, piping layouts, and component configurations. The

calculated travel times are shown in the following table:

Table IPSB-B.1-1 STEAM TRAVEL TIME Steam travel time to crossover piping mid-point (seconds)

OLTP (3,293 MWt)

CLTP (3,458 MWt)

EPU (3,952 MWt) Unit 1 10.70 NA 8.93 Unit 2 10.49 10.12 8.63 Unit 3 10.47 10.10 8.62 The radiological decay of N-16 and C-15 was then calculated for

those travel times, and a fractional increase in the radiation

level was determined for each condition. The results for

operation of all three BFN units were: an approximate 10%

increase from OLTP to CLTP and an approximate 32% increase from

CLTP to EPU.

By applying these increases to the GE dose projection for the

OLTP condition, the annual dose to members of the public offsite

and onsite were determined for CLTP and EPU under normal water

chemistry conditions (i.e., no hydrogen injection). For

calculation purposes, assuming three units at CLTP and EPU, the

annual dose to a member of the public at the nearest terrestrial

site boundary would be 1.2 and 1.5 mrem, respectively. Currently

both Unit 2 and Unit 3 are operating with the addition of Noble

ChemŽ (platinum and rhodium) to the reactor coolant system and

with reduced hydrogen injection in the feedwater for IGSCC

mitigation. For three units at CLTP under reduced hydrogen

injection, these values would be increased by approximately 25%

to 1.4 and 1.9 mrem, respectively. The total dose to a member of

the public includes effluent doses; however, these are negligible

in comparison to the direct and skyshine radiation doses.

E1-4 Therefore, the projected annual doses are well within the 25 mrem dose limit of 40 CFR 190 for an offsite member of the public.

The limiting dose receptors for members of the public would be

those onsite (e.g., food vendors) because their work locations

are nearer to the turbine building. The maximum annual dose to

vendors would not likely exceed 18 mrem under EPU and Noble ChemŽ

water chemistry and reduced hydrogen injection conditions on all

three units. Consequently, the 10 CFR 20.1301 annual dose limit

of 100 mrem for a member of the public onsite would not be

exceeded. Therefore, the projected annual dose to an onsite

member of the public will be well within the dose limit.

Furthermore, following Unit 1 restart, with the reduction in

restart workers and the re-location of a major portion of the

site population into permanent structures farther from the

turbine building, the increase in collective site dose from

direct and skyshine radiation external to the plant structure is

projected to be approximately 10 person-rem per year for three

units at EPU conditions under Noble ChemŽ water chemistry and

reduced hydrogen injection.

NRC Request IPSB-B.8 Section 8.5.1, Normal Operations, of Enclosure 4 of the [June 25, 2004 and June 28, 2004 submittals for BFN Units 2 and 3 and BFN

Unit 1, respectively] submittal states that, due to the

conservative shielding design, the increase in radiation levels

resulting from EPU will not affect the radiation zones for the

various areas of the plant. This appears to be based on an

assumed linear increase in radiation source term with power

level. However, the increase in N-16 activity in the turbine

building is an inverse exponential function with decay time, not

a linear function of reactor power. Verify that the radiation

zoning in all areas containing the steam and feed systems will be

unaffected by EPU.

TVA Reply to IPSB-B.8 Under current licensed thermal power (CLTP) (i.e., 3458 MWt) conditions for BFN Units 2 and 3 with Noble ChemŽ water chemistry

and reduced hydrogen injection in the feedwater for IGSCC

mitigation, all of the steam-affected areas, with the exception

of the reactor feed pump turbine rooms, are locked high radiation

areas (LHRA). This includes the reactor and turbine steam

tunnels, moisture separator rooms, turbine rooms, high and low

pressure heater rooms, condenser rooms, moisture separator drain

pump and tank rooms, steam packing exhauster rooms, steam jet air

ejector rooms, and hydrogen recombiner rooms. The reactor feed

pump turbine rooms are posted as radiation areas at the entrances

with smaller high radiation areas located inside the rooms E1-5 enclosing the turbine and pump areas. The areas on the turbine roof over the turbine rooms are controlled as high radiation

areas. Although BFN Unit 1 is currently not operating, nor is it

currently licensed to operate above 3293 MWt, it is expected that

its radiation levels will be consistent with the radiation levels

of Units 2 and 3.

Under EPU conditions, the radiation levels are conservatively

expected to increase by approximately 32% over the CLTP

conditions. This is based on increased steam flow, reduction in

steam travel time, and reduction in the radiological decay of

N-16 and C-15. However this increase will not be enough to

require changing the radiation area posting at the entrance to

the rooms. In addition, the radiation zoning and posting outside

the steam-affected area rooms are not expected to change due

to EPU.

To ensure that proper postings are maintained, dose rates will be

monitored in these environs during power ascension as part of the

planned EPU testing.

E2-1 ENCLOSURE 2 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

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

SUPPLEMENTS TO RESPONSES TO NRC ROUND 2 REQUESTS FOR ADDITIONAL INFORMATION -

SPLB-A.1, SPLB-A.2 and SPLB-A.3 By letters dated December 19, 2005 (ADAMS Accession Nos.

ML053560194 and ML053560186) TVA submitted responses to NRC

Round 2 requests for additional information (RAIs) regarding

TVA's applications for extended power uprate of BFN Unit 1 and

BFN Units 2 and 3, respectively. As a result of discussions with

the NRC staff, TVA is supplementing its prior responses to RAIs

SPLB-A.1, SPLB-A.2 and SPLB-A.3 regarding fuel pool cooling. The

responses to the subject RAIs are the same for all three BFN

units.

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

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 E2-2 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. The following discussion is

provided to clarify and provide supplemental information on the

BFN spent fuel pool cooling system and is presented in the format (including numbering) of Attachment 2 to Matrix 5 of RS-001, "Review Standard for Extended Power Uprates," Revision 0, December 2003.

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

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 removal (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 E2-3 system is transferred from the condensate storage tank to 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 a n d 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. 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 E2-4 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 li m it. 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.

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 E2-5 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.

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.

E2-6 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.

E2-7 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.

E2-8 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

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.

E2-9 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.

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.

E2-10 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-

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 E2-11 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) (As discussed with the NRC staff on April 13, 2006, TVA will address this item in a supplemental reply.)

(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

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 E2-12 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.

E3-1 ENCLOSURE 3 TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

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

COMMITMENT LISTING Prior to implementing EPU, 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.