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{{#Wiki_filter:}} | {{#Wiki_filter: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 hours 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 hours 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 hours after reactor shutdown and includes 45 hours 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 hours after | |||
reactor shutdown and includes 115 hours 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 hours after reactor shutdown and includes 115 hours 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 hours 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.}} |
Revision as of 12:28, 11 November 2018
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
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
R. G. Jones, POB 2C-BFN
G. V. Little, NAB 1A-C
R. F. Marks, Jr., PAB 1C-BFN
B. J. O'Grady, PAB 1E-BFN
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.