ML061040217
| ML061040217 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 04/13/2006 |
| 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: ML061040217 (22) | |
Text
April 13, 2006
TVA-BFN-TS-418
TVA-BFN-TS-431
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.
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
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
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
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.