ML20236G500
| ML20236G500 | |
| Person / Time | |
|---|---|
| Site: | Callaway  |
| Issue date: | 07/28/1987 |
| From: | Schnell D UNION ELECTRIC CO. |
| To: | NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM) |
| References | |
| ULNRC-1561, NUDOCS 8708040287 | |
| Download: ML20236G500 (14) | |
Text
' Union
.. Etscraic a
1901 Gratiot Street, St. Louis Donald F. SchneII Re President July 28, 1987 U.S. Nuclear Regulatory Commission ATTN:
Document Control Desk Washington, DC 20555 Gen tl emen :
ULNRC-1561 DOCKET NUMBER 50-483 CALL AWAY PLANT RESPONSES TO QUESTIONS ON CYCLE 3 RELOAD USING WESTINGHOUSE VANTAGE 5 FUEL ASSEMBLIES References
- 1) ULNRC-1470 dated March 31, 1987
- 2) ULNRC-1471 dated March 31, 1987
- 3) ULNRC-1491 dated April 15, 1987
- 4) ULNRC-1525 dated June 5,1987
- 5) ULNRC-1535 dated June 18, 1987
- 6) ULNRC-1554 dated July 16, 1987
- 7) NRC Letter dated July 9,1987 f rom T. W.
Alexion to D. F.
Schnell References 1 through 6 transmitted the license application and additional supporting information for Callaway Cycle 3.
Attached to this letter are responses to your request for additional information transmitted by Reference 7.
Reference 2 transmitted our application to increase the licensed power level for Callaway.
Part of the information supplied in Attachment 1 is more directly related to the uprating application and is being provided to allow both the Cycle 3 and uprating reviews to proceed.
If there are any further questions please contact us.
very truly yours, 87080402B7 8707pg DR ADOCK 05000403 p
i PDR Donald F.
Schnell DS/ lad Attachment e
Maihng Address: P.O. Box 149, St. Louis, MO 63166
l l
STATE OF MISSOURI )
)
Donald F.
Schnell, of lawful age, being first duly sworn upon oath says that he is Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof; that he has executed the same for and on behalf of said company with full power and authority to do so; and
)
that the facts therein stated are true and correct to the best of his i
knowledge, information and belief.
l i
By M
Donald F.
Schnell i
Vice President Nuclear SUBSCRIBED and sworn to before me this d M day of
,198[
k lN4 BARGAR J.PFA f40Tf.l!Y fubuC, STATE OF MISSOURI MY COMM SSION EXPlRES APRll 22, 1989 ST. LOUIS COUNTY
I cc:
Gerald Charnoff, Esq.-
Shaw, Pittman, Potts & Trowbridge
)
2300 N. Street, N.W.
Washington, D.C.
20037 Dr. J. O. Cermak CFA, Inc.
4 Professional Drive (Suite 110)
Gaithersburg, MD 20879 W.
L.
Forney Chief, Reactor Project Branch 1 U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, Illinois 60137 Bruce Little Callaway Resident Office U.S. Nuclear Regulatory Commission RR#1 Steedman, Missouri 65077 Tom Alexion (2)
Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Mail Stop 316 7920 Norfolk Avenue Bethesda, MD 20014 Ron Kucera, Deputy Director Department of Natural Resources P.O. Box 176 Jefferson City, MO 65102 Manager, Electric Department Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102 l
l Q____-_____-_____
1.
bec:
3456-0021.6 Nuclear Date DFS/ Chrono D.
F.
Schnell J.
E.
Birk J. F. McLaughlin A.
P. Neuhalfen R.
J. Schukai M.
A.
Stiller G..L. Randolph D.
E.
Shain H. Wuertenbaecher D. W. Capone A. C. Passwater R.
P. Wendling T.
H. McFarland R.
D.
Affolter D.
E.
Shafer D.
J. Walker O.
Maynard (WCNOC)
N.
P. Goel (Bechtel) l G56.37 (CA-460)
[
i Compliance (J. E. Davis)
NSRB (Sandra Auston) 3456-0031.2.l(1057) l i
l
RESPONSES TO QUESTIONS ON CYCLE 3 RELOAD USING WESTINGHOUSE VANTAGE 5 FUEL ASSEMBLIES 1.
Provide a sunmary of the results of the reanalysis which demonstrates that plant operation at the burnup levels associated with V-5 fuel at the two reactor thermal power levels of 3411 MWt and 3565 MWt would not adversely impact the fuel pool cooling capability and fuel building UVAC performance.
The fuel pool results should be in tabulated form identical to those currently in FSAR Table 9.1-4.
Response
New thermal-hydraulic analyses of the spent fuel pool cooling system and fuel building HVAC performance, to suppor t the Callaway uprating. and to assure that depleted VANTAGE-5 assemblies can be stored, have oeen performed.
These analyses assumed a reactor thermal power level of 3565 MWt which bounds the case of.3411 MWt.
Both OFA and. VANTAGE 5 fuel were considered, with VANTAGE 5 being limiting due to higher burnup.
T: = results of the reanalyses on the spent fuel pool heat loads and bulk temperatures are provided in Table 1 for the normal refueling and the full core offload cases of VANTAGE 5.
These analyses assumed the spent fuel would not be moved into the pool for 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> following the time the reactor is subcritical and a nominal time of 1/2 hour to move each assembly from the reactor to the spent fuel pool.
These times are based on physical limitations for plant cooldown, preparing the plant for refueling operations and the time associated with the actual movement of fuel.
The decay heat loads for the equilibrium discharge (84 assemblies) were calculated assuming 35-day refueling outages and a 90% capacity f actor.
The equilibrium discharge is composed of groups of assemblies with similar exposure histories.
The exposure histories are based on limiting predicted region-average burnups (maximum 50,000 MWD /MTU).
The analyses assumed a nominal reload of 84 assemblies.
Although Cycle 3 reload will replace 96 assemblies, this will not have a significant impact because of the relatively few assemblies in the pool at this time and the lower burnups of the assemblies to be stored in the pool compared to the value assumed in the analyses.
The combined effects of the uprating and VANTAGE 5 fuel increase the heat load to the fuel pool and to the fuel building HVAC.
Analysis, however, shows that the fuel building HVAC system has sufficient margin to maintain the l
temperature below 1040F, and the fuel pool temperature is kept below Standard Review Plan limits.
To cool all areas of the building, conditioned outside air is supplied to all the equipment rooms and to the fuel pool area.
The exhaust air from the equipment rooms is drawn through the fuel pool area and is filtered and discharged through the unit vent.
The temperature of the fuel pool area is dependent upon the heat exhausted f rom the equipment rooms, solar, lighting and heat transfer f rom the fuel pool.
The combined effects of the plant uprating, VANTAGE 5 fuel and extended burnup increase the fuel pool temperature from 1200 F. to 1400F.
This results in a greater latent and sensible heat load to the fuel pool area.
Other areas of the fuel building are not appreciably affected by the f uel pool temperature change.
The fuel pool area temperature was recalculated based.on a 140 F pool temperature, summer design conditions and all other loads unchanged.
Analysis shows that the existing HVAC system is capable of maintaining the area at 103 F and 50% relative humidity.
1 I
(.
j
6 t
9V1' i
Wg: l TABLE 1 FUEL POOL COOLING AND CLEANUP SYSTEM DESIGN PARAMETERS 1.
Spent fuel pool.(SFP) capacity, assemblies 1344
' 2.
SFP pool water volume, gal 400,000 (level at 23 feet above top of racks) 3.
Boron concentration of the SFP water, ppm 2000
/
7<
\\
4a. Heat Loads and Bulk Pool Temperatures (Normal Refueling)
Refueling Decay Total Heat Bulk Pool Number Time (hrs)
Load (MBTU/HR).
Temperature (9F)
' l 192 18.06 128.9 2
12228 18.92 130.1 3
24263 19.27 130.6 4
36299 19.49 130.8 5
48335 19.67 131.1 6
60371 19.83 131.3 7
72406 19.99 131.5 8
84442 20.14 131.7 9
96478 20.29 131.9 10 108513 20.43 132.1 11 120549 20.57 132.3 12 132585 20.70 132.4 13 144621 20.83 132.6 14 156656 20.95 132.8 1
l L__ _ _ _ _ __ _
. (L I
TABLE 1 (Sheet 2) l 4b.
Heat Loads and Bulk Pool Temperatures (Full Core Offload)
Refueling' Decay Total Heat Bulk Pool Number Time (hrs)
Load (:MBTU/HR)
Temperature (OF)
Offload 246.5 36.41 153.3 1
12282 37.26 154.4 2
24318 37.61 154.9 3
36354 37.83 155.2 i
4 48389 38.01 155.4 5
60425 38.17 155.6 6
72461 38.33 155.8 l
7 84497 38.48 156.0 i
8 96532 38.62 156.2 9
108568 38.76 156.4 10 120604 38.90 156.6 11' 132639 39.03 156.8 j
12 144675 39.16 156.9 t
13 156711
-39.29 157.1 i
14 168747 39.40 157.2 Notes:
(1)
For computation of thermal parameters, the SFP is considered isolated f rom the fuel transfer canal and 1
the cask loading pool.
(2) 84 assemblies are assuaed to be discharged at each refueling.
(3)
Bulk pool temperatures are based on the use of one (of two) 100% cooling water train with 105 F CCW.
l (4)
Residual decay heat is based on NRC BTP ASB 9-2.
( 5)
MBPH = 10 ** 6 Btu /hr I
_m-._____J
. Provide the overall thermal conductance (BTU /SqFt-Hr PF) for 2.
each spent fuel pool cooling system heat exchanger.
Response
Union Electric performed the spent fuel pool (SFP) heat exchanger performance calculations using the heat exchange ef festiveness method of Kays and London (Compact Heat Exch ange rs, 1958).
A separate calculation was performed to confirm the SFP heat exchanger data supplied by the manufacturer.
That confirmatory calculation yielded values for the overall thermal conductance which dif fered by less than 2% from the manufacturer-supplied value of 337 BTU /SqFt-Hr-F.
b
... 3.
Provide the revised Component Cooling Water System (CCWS) heat load duties.(in a tabulated form) with the plant operating at 3411 MWt and 3565 MWt reactor core power levels using V-5 fuel.
Verify that the CCWS can adequately cool the safety related systems and components it serves, including the spent fuel pool cooling system.
RESPONSE
Table 2 contains the current calculated heat loads for the CCW heat exchangers.
As assumed therein, Train A handles one RHR hear exchanger, and various other loads, while Train D mainly handles the second RHR heat exchanger.
Table 3 contains the revised heat loads.
The only changes are increased heat loads for the RHR heat exchangers and for the fuel pool cooling heat exchanger.
Operation at the uprated power level and use of VANTAGE 5 fuel have insignificant impact on the remaining heat loads.
For the Normal case, the increased fuel pool cooling heat exchanger load can be handled by the CCW heat exchanger with minimal change to operating' conditions.
The design case for a CCW heat exchanger shows a secondary (tube) side temperature rise of ll.40F with a duty capacity of 77.18 MBTU/HR (see FSAR Table 9.2-10).
A total load of 90.63 MBTU/HR (from Table 3) represents a 2 F increase in tube side exit temperature (other factors being equal) from design.
The actual heat loads on the CCW heat exchanger are less than those assumed in design.
Table 4 presents results from data taken on 7/18/86.
For the given components, calculated values indicate a total load of 48.4 MBTU/HR, whereas the actual total load was 38.7 MPTU/HR, or 20% lower.
Furthermore, the actual totai load was significantly lower than the design case duty capacity of one CCW heat exchanger (almost half of the 77.18 MBTU/HR).
This reduction is due partly to the f act that not all of the nonessential components (those below the dashed line in Tables 2 and 3) are in operation at the same time.
The :?6nction is also due to the conservatively large heat
'i load assumea for the fuel pool cooling heat exchangers.
This heat load represents a full fuel pool with freshly discharged assemblies and high burnups.
Actual heat loads would be reduced (due to decay) the majority of the time.
For the '1hutdown (at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />) case, the increased RHR heat exchanget heat load results in an increased time to cool down.
Yne original 16 hour1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> plant design cooldown time from l
350*F to 140 F will now require 19.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> (reference ULNRC-0 l
1471 dated 3/31/87).
This increase in plant cooldown time is not considered to be significant; safety requirements are I
l
. still satisfied with respect to a single train cooldown under accident conditions.
For the Post-LOCA case, the revised 195 MBTU/HR (no previous comparable calculation) represents a conservative upper bound on the heat load based on a conservatively high sump water temperature.
This heat load was determined by the CCW flowrate, service water flowrate and temperature, the recirculation flowrate, and the sump water temperature.
With the exception of the sump water temperature, all of these parameters remain unaffected by the uprated power level.
Sump water temperature could increase by some small amount due to the higher initial reactor power level.
Like the shutdown cooldown time, a higher post-LOCA heat load equates to a longer time for cooldown.
This potential increase in cooldown time is not considered to be significant.
Based on the above discussion, the CCW system will adequately cool the involved safety related systems and compone nts, including the spent fuel pool cooling system, at the 3411 MWt and 3565 MWt reactor core power levels with the use of VANTAGE-5 fuel.
^*
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d ' TABLE __2 3
{
t CCW HX Train A Heat Loads _(MBTU/HR) l Shutdown I
Component Normal (0 4 hrs)
Post-LOCA RHR UX 0
116 Not Calc.
RHR Pump Seal Cooler 0.03 0.03 0.03 CCP Bearing Oil Cooler 0.08 0.08 0.08 SI Bearing 011 Cooler 0
0 0.06 Fuel Pool Cooling HX 16.89 0
0 Excess Letdown HX 0
0 0
RCP Coolers 9.28 2.32 0
2.24 0
RCDT EX 2.24 i
Letdown HX 16.0 16.0 0
Seal Water HX 2.40 2.40 0
Recycle Evaporator Package 8.81 8.81 0
Waste Gas Compressors 0.25 0.25 0
PDP Oil Cooler 2.60 2.60 0
Aux. Steam Rad Monitor 0.23 0.23 0
PASS Sample Coolers 0.10 0.10 0.10 0.30 0
-Catalytic Hydrogen Recombiners 0.30 O.59 0
NSS-Sample Cooler 0.59 Waste Evaporator Package 8.81 8.81 0
Secondary Waste Evap. Pack.
17.8 0
0 Reverse Osmosis Unit 0.16 0.16 0
Total 86.57 160.92 Not Calc.
CCW HX Train B Heat Loads (MBTU/HR)
Shutdown Component Normal
(@ 4 hrs)
116 Not Calc.
RHR Pump Seal Cooler 0
0.03 0.03 CCP Bearing Oil Cooler 0
0.08 i.08 SI Bearing Oil Cooler 0
0 0.06 Total 0
116 11 Not Calc.
l l
-9 TABLE 3 REVISED CCW IIX LOAD _S CCW HX _Tr ain A Heat Loads _(MBTU/HR)
Shutdown
'1 Normal
( @ l4 h r s)
Post-LOCA Component 0
(118
)
(195
)
RHR HX RHR Pump Seal Cooler 0.03 0.03 0.03 CCP Bearing Oil Cooler 0.08 0.08 0.08 SI Bearing 011 Cooler 0
0 0.06 Fuel Pool Cooling HX (20.95) 0 0
Excess Letdown HX 0
0 0
RCP Coolers 9.28 2.32 0
0 RCDT HX 2.24 2.24 Letdown HX 16.0 16.0 0
2.40 0
Seal Water HX 2.40 i
Recycle Evaporator Package 8.81 8.81 0
Waste Gas Compressors 0.25 0.25 0
PDP Oil Cooler 2.60 2.60 0
Aux. Steam Rad Monitor 0.23 0.23 0
PASS Sample Coolers 0.10 0.10 0.10 Catalytic Hydrogen Recombiners 0.30 0.30 0
NSS Sample Cooler 0.59 O.59 0
8.81 0
Waste Evaporator Package 8.81 0
0 Secondary Waste Evap. Pack.
17.8 Reverse Osmosis Unit 0.16 0.16 0
Total (90.63)
(162.92)
(195.27)
CCW HX Train B Heat Loads (MBTU/_HR)
Shutdown Normal
(@ 4 hrs)
Post-LOCA Component 0
(118
)
(195
)
0.03 0.03 CCP Bearing Oil Cooler 0
0.08 0.08 SI Bearing 011 Cooler 0
0 0.06 Total 0
(118.11)
(195.17)
(
) represent values changed from Table 2.
t
.e Table 4 CCW HX Heat Loads on 7/18/86 Components in Operation:
CCP Bearing Oil Cooler Fuel Pool Cooling HX RCP Coolers RCDT HX Letdown HX Seal Water HX Waste Gas Compressor Catalytic Hydrogen Recombiner Secondary Waste Evap.. Pack Design Heat Load Per Table 1:
Total = 48.4 MBTU/HR
- Actual Heat Load:
Total = 3 8.7 MBTU/HR
- Heat load f rom fuel pool cooling HX assumed to be zero.
)
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