ML20090K368
| ML20090K368 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 12/08/1983 |
| From: | Snow A DUKE POWER CO. |
| To: | |
| References | |
| A-091, A-91, NUDOCS 8405240182 | |
| Download: ML20090K368 (23) | |
Text
_ _ - _ - - _
[
i-q Icc=ra t.
A.
UNITED STATES OF AMERICA D
i (V
NUCLEAR REGULATORY COMMISSIO -j ??;AR 201934201934a _
V\\
DocKrmIGS
'/l BEFORE THE ATOMIC SAFETY AND LICENSIN }OAaggC1 4
D In the Matter of
)
Ib
)
DUKE POWER COMPANY, et al.
)
Docket Nos. 50-413
)
50-414 (Catawba Nuclear Station,-
)
Units 1 and 2)
)
TESTIMONY OF A. LOWELL SNOW 1
Q.
PLEASE STATE YOUR NAME.
2 A.
Lowell Snow 3
Q.
BY WHOM ARE YOU EMPLOYED AND IN WHAT CAPACITY?
4 A.
I am a Design Engineer II employed by Duke Power Company, 422 5
S. Church Street, Charlotte, North Carolina 28242.
A statement of 6
my qualifications is attached to this testimony as Attachment B.
7 Q.
DESCRIBE THE AREAS OF YOUR RESPONSIBILITIES AS A DESIGN 8
ENGINEER II.
9 A.
My job responsibilities include spent fuel decay heat and criticality 10 evaluations.
Prior to my present job I was responsible for Catawba 11 fluid systems design.
12 Q.
.WHAT IS THE PURPOSE OF YOUR TESTIMONY?
13 A.
The purpose of my testimony. is to address those portions of 14 Contention 16 dealing with. (1) the ability of the ' spent fuel pool 15 cooling system to maintain anticipated pool water temperatures at or 16
. below the NRC's acceptance criteria with Oconee and McGuire,. as 17 well as Catawba, spent fuel; and (2) the criticality aspects of the 18 expanded fuel pools.
O MPEMPau"@
O
~
1 COOLING SYSTEM 2'
Q.
DESCRIBE THE PRESENT SPENT FUEL POOL ~ COOLING SYSTEM.
3 A.-
The spent fue' pool cooling system consists of -two cooling loops, 4
one purification loop and one skimmer loop.
The pool cooling 5
subsystem consists of two full capacity pumps, each of which are 6
designed to pump 2840 gallons / minute, two full capacity heat 7
-exchangers, each of which are designed to maintain the spent fuel 8
pool temperatures below 150 F under normal heat load conditions (as 9
described on p.
3, line 19), and associated piping and valves 10 sufficient to take suction from the pool and return the cooled water 11 to the pool.
This equipment is arranged in two loops (sometimes 12 referred to as trains), each with one pump, one heat exchanger 13 and associated piping and valves.
The details of this system are I
14 set forth in FSAR Section 9.1.3.
15 The pool purification subsystem consists of a fuel pool cooling 16 pre-filter, fuel pool cooling demineralizer and fuel pool cooling 17 post-filter.
The spent fuel pool skimmer loop consists of a skimmer 18 trough, strainer, skimmer pump, and filter.
19 Q.
HOW DOES THE PRESENT DESIGN OF THE SPENT FUEL POOL 20 COOLING SYSTEM COMPARE TO THE ORIGINAL DESIGN?
21 A.
The design is essentially -the same. Due to the increased length of 22 the fuel pool the -return lines from the spent fuel pool cooling heat 23-exchangers were lengthened.
The spent fuel pool cooling - pump 24 internals were also modified to increase the flow rate through the 25 spent fuel pool cleanup filters and demineralizers so that one full 26 pool volume per day could still pass through the purification loop.,
l
1-The spent fuel cleanup filters and demineralizers were re-sized to 2
accommodate the additional flow.
-3 Q.
IS THE PRESENT SPENT FUEL POOL COOLING SYSTEM ADEQUATE
'4 TO PROVIDE SUFFICIENT SPENT FUEL POOL COOLING?
5 A.
Yes.
Consistent with guidance found in American National 6
Standards Institute (ANSI) N210-1976 " Design Objectives for Light 7
Water Reactor Spent Fuel Storage Facilities at Nuclear Power 8
, Stations," the present spent fuel pool cooling system is designed to 9
maintain the temperature of the spent fuel pool below 150 F.
(This 10 assumes that both pool cooling trains are operating under abnormal 11 heat load conditions.) The present spent fuel pool cooling system 12 also complies with more recent NRC guidance found in Standard 13 Review Plan 9.1.3, " Spent Fuel Pool Cooling and Cleanup System,"
Q 14 and Regulatory Guide 1.13, " Spent Fuel Storage Facility Design 15 Basis. "
In addition, it should be noted that the spent fuel pool 16 cooling system is capable of maintaining the temperature of the
-17 spent fuel pool 'below boiling in all cases, normal and abnormal, 18 with the loss of one cooling train.
19 Q.
DEFINE NORMAL HEAT LOAD AND ABNORMAL HEAT LOAD 20 CONDITIONS.
21 A.
As shown.in Figure 1 (attached), normal heat load is the postulated
'22 maximum - achievable heat generation rate (Btu /Hr) resulting from
-23 the storage of spent fuel in' a Catawba spent fuel pool which 24 maintains a full core reserve.
25 The abnormal heat load is ' the postulated maximum achievable
'26 heat generation ' rate resulting from the storage of spent' fuel in a 27 spent fuel pool following a full core discharge.
~,
ix..
1 Q.
GIVEN THE INCREASED SIZE OF-THE SPENT FUEL POOL WHY
.J' 2
WERE NOT MODIFICATIONS REQUIRED TO SATISFY THE COOLING 3
SYSTEM DESIGN CRITERIA?
4 A.
Calculations indicated that no change in the cooling system was 5
necessary and that the heat loads of the expanded pool could be 6
met with the. originally designed cooling system.
This is due 7
primarily to the greatly reduced heat output of spent fuel as it 8
decays with time.
For example, the heat output from a spent fuel 9
assembly decreases by a factor of 10 during the first year of 10 decay.
It should be noted that -an increase in pump flow rate for 11 purposes of the purification loop was made to accommodate the 12 larger volume of water necessary for purification.
However, this 13 increased flow rate is not included in the flow to heat exchangers 14 for cooling purposes in cooling response calculations.
15 Q.
DESCRIBE THE COOLING RESPONSE CALCULATIONS.
16 A.
The spent fuel pool temperatures that result from the assumed heat l
17 loads described in Figure 1 are graphically presented in Figure 2.
18 These temperatures are calculated by using heat transfer equations 19 for heat exchanger evaluations using the effectiveness method found 20 in standard textbooks.
(See Attachment A).
The calculations are
{'
21 conservative in that they assume only one mechanism for heat l
22 transfer, which is heat exchanger cooling.
They also assume that 23 heat exchanger cooling water enters heat exchangers at the 24 maximum - temperature of. 100 F.
No credit is taken in these 25
. calculations.for convection', evaporation, radiation heat transfer to l
l 26
.the fuel pool surroundings, conduction through.the pool walls to l t
. _,,. ~,
~ - -
T 1
the environment, or. heat load decay in the spent fuel pool during 2
the heatup times to equilibrium conditions.
3 Q.
HOW-DO YOU DETERMINE NORMAL AND ABNORMAL HEAT LOADS?
4 A.
As set forth in Figure 1, the heat loads for various conditions are 5
determined using ' the NRC decay heat curve (Branch Technical 6
Position ASB 9-2) as set forth in S.R.P. 9.1.3.
The 7 day block 7
under the three normal cases described in Figure 1 results-from a 8
refueling (i.e.
61 Catawba spent fuel assemblies being transferred 9
to the spent fuel pool all being decayed for 7 days, which is a 10 conservative estimate of the time required to accomplish the transfer j
j 11 of the fuel). The remainder of the normal design heat load and the 12 normal expanded Catawba-only heat load are comprised of additional 13 yearly spent fuel discharges such that all available storage spaces 14 (excluding full core discharge reserve) are filled with Catawba fuel 15 from previous refuelings.
This results in an increase in heat load 16 from 13.8 million BTU /HR for the normal fuel pool design to 17 17
. million BTU /HR for the normal expanded case.
18 For the normal expanded combined case (that is, Catawba, 19 Oconee and McGuire spent fuel in the Catawba spent fuel pool),
20 heat load calculations assume an increase due to other units' (eg, 21 McGuire and Oconee) spent fuel being added earlier in the decay l
22 sequences than was assumed for the - Catawba-only sequences.
23 Spent fuel from these stations is assumed to be decayed for 270 24
. days, which 'is conservative when compared to the stipulated 5-year 25 old spent fuel.
It should be noted that if 5-year old spent fuel is 26' assumed, the resulting value will be 17.3 million BTUs/ hour rather h
27 than the 20.6 million BTUs/ hour shown in Figure l'.
1 For the three abnormal cases shown in Figure 1, the 7 day 2
block now results from a full core discharge (FCD) of 193 3
assemblies which have been decayed for 7 days. The 25 day block 4
results from an immediate past refueling outage that has decayed 5
for 25 days.
Twenty-five days decay for a refueling batch (with 6
an assumed full core discharge) was selected because an evaluation 7
of decay heat curves reflects that an assumed 11 day irradiation 8
time for the full core prior to discharge approximates the maximum 9
combined heat load of the full core discharge plus the immediate 10 past refueling outage.
11 The remainder of the abnormal heat load for both the current 12 and earlier fuel pool designs is comprised of additional yearly spent 13 fuel discharges such that all available storage spaces are filled with 14 Catawba fuel.
As shown in Figure 1, this results in an increase in 15 heat load from 35.9 million BTU's to 39.0 million BTU /HR.
16 For the abnormal expanded combined case (that is, Catawba, 17 Oconee and McGuire spent fuel pool), heat load calculations assume 18 the 7 day and 25 day blocks discussed immediately above.
They 19 also assume an increase due to other units' ( e.g., McGuire and 20 Oconee) spent fuel being added earlier in the decay sequences than 21 was assumed for the Catawba-only sequences.
Spent fuel from i
22 Oconee and McGuire is assumed to be decayed for 270 days, which l
23 is conservative when compared to the stipulated 5-year old spent 24 fuel.
It should be noted that if 5-year old spent fuel is assumed, 25 the resulting value will be 39.4 million BTUs/ hour rather than the i
l 26 42.7 million BTUs/ hour shown in Figure 1.
O,
-4 s
m,
1 It should be noted that these heat load calculations are 2
conservative in that they assume (1) that only 7 days is required 3
to refuel the reactor; (2) that the reactor units operate 4
continuously (except during assumed 7 day refueling outage); and 5
(3) that there is maximum burnup of all discharged fuel.
In 6
. addition, the combined heat load cases assume that the freshest 7
(>270 days) spent fuel available at the McGuire and Oconee sites is 8
transhipped to Catawba rather than the older, cooler, spent fuel 9
available at these sites.
10 Q.
DESCRIBE THE SPENT ' FUEL POOL TEMPERATURES THAT YOU 11 HAVE CALCULATED.
12 A.
Figure 2 (attached) illustrates the results of taking six of the heat i
13 load values presented in Figure 1 and evaluating each of them first 14 with one cooling train in operation and then with two cooling trains 15 in operation.
For each of the normal cases it is observed that with 16 one cooling train available the resulting maximum fuel pool 17 temperature is less than 140 F.
As would be expected, the 18 increased heat load associated with the expanded pool results in 19 temperatures higher than the original design, but in all cases well 20 below the design basis.
Since the design basis is still met, these 21 changes in temperature do not represent a decrease in safety 22
. margin, nor have they resulted in a change of fuel ' pool cooling 23 system capacity.
24 With respect to the abnormal cases,- operation with two cooling 25 trains results in maximum fuel pool temperatures below 150 F.
(The 26 two cooling train assumption is consistent with the above-referenced -
h 27 Standard Review Plan and Regulatory Guide).
In addition, the.-
1-abnormal cases with only one cooling train in operation are shown to
'2.
be well below 212 F (boiling).
Since this additional design 3
consideration is met, these changes in temperature do not represent 4
a decrease in safety margin, nor have they resulted in a change of 5
fuel pool cooling system capacity.
I 6
Q.
GIVEN THE INCREASE IN THE STORAGE CAPABILITY OF THE 7
EXPANDED FUEL. POOL, WHY IS THE TEMPERATURE INCREASE 8
RELATIVELY LOW?
9 A.
The calculated design temperature of the fuel pool is a function of 10 the decay heat load and the performance of the spent fuel pool 11 cooling heat exchangers. Equilibrium conditions are assumed in fuel 3
12 pool temperature calculations (i.e., for an assumed constant heat 13
- load, a constant fuel pool temperature is calculated).
At Q
14 equilibrium,; the spent fuel pool cooling heat exchangers are 15 rejecting an amount of heat equal to the spent fuel pool heat load.
1 16 When additional heat load is added to the spent fuel pool, the 17 temperature of the pool rises, resulting in additional heat being i
18 transferred across the heat exchangers.
(This assumes constant 19 cooling water temperature and heat exchanger flow rates).
A new 20 equilibrium temperature is established when the heat exchanger heat 21 rejection again equals the fuel pool heat load. The relative increase f
22 in the calculated equilibrium temperature, however, is not equal to 23 the relative increase in heat load since heat exchangers reject heat i
24 more effectively-at increased " hot side" temperatures _(assuming all i
25 other parameters are constant).
26 As observed in Figure.1, the design heat loads for -the O
27 expanded pools did not increase proportienately with the pool l.
i l I t
II,
1 volume increase due to the ' dominating effect of assumed fresh 2
Catawba discharges.
This fact, coupled with the increased heat 3
exchanger heat rejection capability at increased fuel pool 4
temperatures, results in calculated pool temperatures for the 5
expanded fuel pool design only slightly increased over the original 6
design.
7 Q.
HAVE YOU ANALYZED THE CONSEQUENCES OF FAILURE OF BOTH 8
COOLING TRAINS, ASSUMING THAT NO MAKEUP WATER IS 9
SUPPLIED AND THE MAXIMUM DECAY HEAT PRODUCTION RATE IS 10 UTILIZED?
11 A.
Yes.
12 Q.
PLEASE EXPLAIN YOUR ANALYSES AND RESULTS.
13 A.
In Figure 3 (attached), the previously described decay heat loads 14 are evaluated to determine the minimum time for onset of boiling and
]
15 the minimum time for assembly uncovery.
The evaluation assumes 16 conservative heat transfer mechanisms of maximum heat generation 17 and no heat losses, such that all of the heat is transferred into 18 increasing the temperature of the water.
At the onset of boiling, it 1
4 19 is assumed that all of the heat generated is utilized in evaporating 20 the water.
Standard heat transfer equations, which are set forth l
l 21 in Attachment A, were used to develop these minimum times.
The 1
22 results of these calculations reflect that for the normal expanded 23 heat load case, at least 25.3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> will elapse before the onset of 24 boiling, and moreover, that at least 150 hours0.00174 days <br />0.0417 hours <br />2.480159e-4 weeks <br />5.7075e-5 months <br /> will elapse before 25 assembly uncovery, assuming no cooling or makeup. With regard to l
26 the normal expanded combined mode, about 21.0 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> will elapse l O 22 before the enset of bo111ng and aseut 138 hours0.0016 days <br />0.0383 hours <br />2.281746e-4 weeks <br />5.2509e-5 months <br /> wi11 e1 apse prior to - - -.
1 assembly uncovery.
With regard to the abnormal expanded mode, 2
the values are 9.8 and 116 hours0.00134 days <br />0.0322 hours <br />1.917989e-4 weeks <br />4.4138e-5 months <br />, respectively, and with regard to 3
the abnormal expanded combined mode the results are 8.9 hours1.041667e-4 days <br />0.0025 hours <br />1.488095e-5 weeks <br />3.4245e-6 months <br /> 4
until the onset of boiling and 106 hours0.00123 days <br />0.0294 hours <br />1.752645e-4 weeks <br />4.0333e-5 months <br /> until assembly uncovery.
5 Q.
GIVEN THE ABOVE TIMES TO ASSEMBLY UNCOVERY, CAN 6
ACTIONS BE TAKEN AT CATAWBA TO PROVIDE REASONABLE 7
ASSURANCE THAT ASSEMBLY UNCOVERY DOES NOT OCCUR?
8 A.
Yes.
There are several sources of water that can provide makeup 9
to the spent fuel pool.
These sources include the Refueling Water 10 Storage Tank (FWST), which can be used for normal makeup 11 requirements, and the assured fuel pool makeup source of Nuclear 12 Service Water (lake water), which provides fully redundant makeup 13 in the unlikely event of loss of all cooling capacity.
Both of the 14 above referenced sources are classified as Nuclear Safety Related 15 sources of makeup.
The 395,000 gallon Refueling Water Storage 16 Tank can provide a gravity fed source of borated water at a rate 17 sufficient to maintain the fuel pool water level during a postulated 18 boiling event to the extent of the tank volume.
19 The assured fuel pool makeup source can provide virtually 20 unlimited makeup to the spent fuel pool at a rate well in excess of 21 the maximum expected boiloff rate, which is less than 100 gallons 22 per minute (gpm).
Also, a half full Storage Tank can provide 23 makeup at approximately 370 gpm and the assured makeup source 24 can provide makeup at approximately 500 gpm from each of two 25 independent redundant trains.
The 500 gallons / minute is available 26 from the Catawba River, which flows at an average rate of $1.97 AU n
-t 1
million gpm.
It is my understanding that these sources can be called upon well within the time calculated to assembly uncovery.
~
2 3
CRITICALITY 4
Q.
HAVE YOU PERFORMED A DESIGN SPECIFIC CALCULATION OF 5
THE REACTIVITY OF THE EXPANDED CATAWBA SPENT FUEL 6
POOL?
7 A.
Yes.
8 Q.
PLEASE EXPLAIN YOUR CALCULATION AND RESULTS.
9 A.
As set forth in FSAR Section 9.1.2.3.1, two criticality analyses 10 were performed, one for Catawba /McGuire spent fuel (which are 11 identical),
and one for Oconee spent fuel.
For the 12 Catawba /McGuire spent fuel case, the assumptions are:
(1) an 13 initial enrichment of 3.5 weight percent U ss (this assumes no 2
14 credit for boron concentration); (2) infinite storage arrays in 15 lateral directions to establish the " worst case" K,gg for the storage 16 rack configuration; and (3) 13k" center to center spacing.
For the 17 Oconee analysis, the assumptions were identical except that the 18 initial enrichment of 3.3 weight percent U23s was assumed.
The 19 methodology outlined in Standard Review Plan 9.1.2, " Spent Fuel 20 Storage"; ANSI N210,
" Design Objectives for LWR Spent Fuel 21 Storage Facilities at Nuclear Power Stations"; and ANSI N18.2, 22 "Nucleer Safety Criteria for the Design of Stationary PWR Plants" 23 was used.
This methodology was used for both the 24 Catawba /McGuire and the Oconee cases.
It should be noted that 25 this methodology directs consideration of:
26 a.-
Accidental tipping, falling or dropping of a spent fuel O
27 assembiFi 1-b.
Accidental tipping, falling or sliding of storage racks during 2
fuel transfer or during seismic events; 3
c.
Stuck fuel assembly / crane uplift forces, 4
d.
Objects that may fall on stored fuel assemblies.
5 The analyses considered these accident situations.
In each of 6
these accident situations, the normal storage configurations (*13.5" 7
center-to-center spacing in racks) with the assumed infinite array 8
and assumed lack of boron yield the highest K,gg values.
The
{
9 calculated worst case K,gg values for these assumptions of each
)
10 specific fuel type are:
l 11 Catawba and McGuire fuel - K,fg = 0.922 12 Oconee fuel - K,gg = 0.915 13 These' values demonstrate design compliance with the referenced I
14 NRC criteria by maintaining a storage array neutron multiplication 7
15 factor (K,gg)
_ 0.95 under all credible normal and accident 16 conditions.
17 Q.
ARE THERE UNCERTAINTIES ASSOCIATED WITH THESE VALUES 7 i
18 A.
Yes.
19 Q.
PLEASE EXPLAIN.
20 A.
Figure 4 (attached) is a graphical representation of the calculated 21 K,gg values.
Figure 4 illustrates how calculational and geometric l
22 uncertainties are treated in a conservative manner.
L 23 For the Catawba and McGuire reactivity calculation, the 24 K,gg = 0.922 value represents the calculated worst case K,gg 25 including consideration of uncertainties.
The calculational 26 uncertainties contribute 0.024 to this K,gg value. The calculational 4
Q 27 uncertainties (determined by benchmark and statistical studies) are <
1 applied conservatively by assuming they occur in the positive
.g D
2 reactivity direction.
The assumption that storage rack dimensional 3
tolerances and assembly positioning in the rack (i.e., geometric 4
uncertainties) occur in the worst possible combination contributes 5
0.014 K,ff to the 0.922 value.
This is conservative since 6
dimensional tolerances and assembly positioning do not occur in the 7
worst combinations, but tend to cancel out over the storage array.
8 Figure 4 also illustrates how the conservative assumptions of 9
infinite array and of no boron affect the calculated K,ff value.
10 The infinite array assumption contributes *0.01 to the total 0.922 11 K
value.
The assumed absence of boron (the normal boron eff 12 concentration will be 2,000 ppm) in the pool contributes s0.20 to 13 the total 0.922 K value.
The K f r a Catawba storage array eff eff 14 containing new (unirradiated) fuel from McGuire or Catawba, 15 therefore, is on the order of 0.674 without the above uncertainties 16 or conservatisms.
17 The Oconee reactivity calculation worst case K value is eff 18 0.915.
The calculational method uncertainties contribute 0.028 K,ff 19 to this value.
The contribution due to worst case treatment of 20 dimension tolerances and assembly positioning has not been 21 separately evaluated for Oconee fuel, but is included in the present 22 results.
The reactivity contributions resulting from the assumed 23 infinite array and assumed lack of boron are 0.01 and 0.20, 24 respectively.
The reactivity value for the storage of new Oconee i
25 fuel in the Catawba storage racks is therefore on the order of 0.677 l
26 without the above uncertainties or conservatisms but assuming worst C
27 case geometry.,
e-y
.,,_-o-w,
,-wy,-
m
~.-
\\
1 It should be noted that additional conservatism exists in the 2-reactivity calculations in that new fuel enrichments are assumed.
)
3 Spent fuel is depleted in fissile U-235 and represents a much less 4
reactive configuration in storage arrays than that assumed in the 5
calculations.
6 Q.
DO THESE VALUES PROVIDE REASONABLE ASSURANCE THAT THE 7
PUBLIC HEALTH AND SAFETY IS PROTECTED?
8 A.
Yes.
They are below the values suggested in above-referenced 9
NRC regulatory guidance which establishes 0.95 as a maximum K,gg 10 value.
11 Q.
WHAT IS SIGNIFICANT ABOUT THE 0.95 K EFFECTIVE VALUE?
12 A.
There is an arbitrary factor of safety between 0.95 and 1.00 K,gg.
13 It should be noted that criticality does not occur until 1.00 K,gg is 14' reached.
Accordingly, any value below 1.00 K,gg is suberitical.
15 Q.
WHAT STEPS HAVE BEEN TAKEN IN THE DESIGN OF THE SPENT 16 FUEL POOL COOLING SYSTEM TO KEEP WORKER EXPOSURE AS 17 LOW AS REASONABLY ACHIEVABLE (ALARA)?
18 A.
Steps taken to reduce worker exposure include the following.
19 First, the depth of the water in the spent fuel pools provides 20 shielding.
Also, the provision for filtration and demineralization 21 removes fission products.
Third, the spent fuel pool heating, i
22 ventilation and air-conditioning system removes radioactive t
{
23 particulates in the spent fuel pool area.
Fourth, the layout of the 24 piping and components in the spent fuel pool cooling and cleaning 25 system is such that worker exposure is controlled. Fifth, radiation f
26 monitors are. provided in the spent fuel ~ pool area to alert plant O:
27 personnel to potential radiation hazards.
i i [
1 I. hercby certify that I have read and understand this document and 2
believe it to be my true, accurate and complete testimony.
3 5
6 A. LoweH Snow' 7
8 9
Sworn to and subscribed before me 10 this 3o day of September,1983.
11 OZd7?/&
14 Notary Public 15 16 Commission Expires 7-/4 -47 Io l
)
i
'O
) _
V
~
FIGURE 1 - HEAT LOAD CASES Pool Filled by Previous Refuelings NORMAL CASES 20.61 Expanded Design 7 day decay combined batch W/1418 spaces-FCD 16 96 Expanded Design 7 day decay Catawba Only batch W/1418 spaces-FCD 13.85 d
W/664 p es-FCD atc$ Y Pool Filled by Previous Refuelings ABNORMAL CASES 42.67 Expanded Design Combined 7 da ecay 25d W/1418 spaces 39.02 Expanded Design 7 day decay 25d Catawba Only FCD W/1418 spaces 35 91 Design 7 day decay 25d W/664 spaces FCD 32.96 P5AR Stage 7 day decay 25d 281 spaces FCD 7
10 20 30 40 50 60 6
HEAT LOAD (BTU /HR X 10 )
4 I
)
FIGURE 2 - FUEL POOL C0 h TEMPERATURE CASES 140*F 212*F NORMAL CASES l
1 l
l 117'F 20.61* - Expanded
\\\\\\\\N Combined l
l g.
l l
114*F l
16.96 - Expanded
\\\\gg Catawba On1Y 128'Fl l
I I
l 112*F l
l 13.85 - Design
\\\\\\]
g W/664 spaces 124*F I
l l
l ABNORMAL CASES ns*r!
42.67 - Expanded
.\\\\\\\\ D D N j73 7 Combined g
I 133jF 39.02 - Expanded
\\\\\\\\\\\\\\\\N l
l 166*F Catawba Only g
l l
130*Fl
\\\\\\\\\\N l
159'F 35.91 - Design V/664 spaces g
I l
l l
l 100 120 140 160 180 200 220 TEMPERATURE (*F)
- (x 106 BTU /HR) l\\\\\\\\\\l Two Cooling Trains l
l One Cooling Train
/'
s FIGURE 3 - LOSS OF CO
.NG/ MAKEUP CASES Time to initiate Boiling NORMAL CASES 138 Hrs 20.61* - Combined 21.0 Storage Hrs 155 Hrs 16.96 - Catawba only Storage +
25.3 Hrs ABNORMAL CASES 106 Hrs 42.67 - Combined g,g Storage Hrs 116 Hrs 39.02 - Catawba only Storage 9.8 Hrs d0 12'O l'80 TIME (HRS.)
- (x106 BTU /HR)
+ FSAR Section 9 1.3.3.1 Reference Case
lll il l
eu la V
y l
r a
o c
t i
a t
lu ir g
C e
R 00 C
0 R
N 1
0 59 0
S NO I
TA 5 3 7
LU 1
9 7
9 1 C
6 L
A 0 8 0
C l
Y e
u F
e T
e C
n A
o E
c R
O 4
ER U
28t 4
2 9 7
9 8. U 6
G I
F 0 08 0
leu
&F ae brwi au tG ac CM 0
0 0
0 0
0 0
0 0
0 0
0 0
9 8
7 6
5 1
0 0
0 0
0 f
fe K
l t
t
Thermodynamic and Heat Transfer Equations Attachment A O
V Basic Mass. Energy and Temperature Relationship for Adiabatic Heatup
- 1) Q = MCpAT whe re: Q = Energy gained or lost from system (Btu).
M = Mass of water in system (1bm).
Cp = Specific heat of water (BtuI bm
- F)
- AT = The total change in temperature of a system of mass M resulting from the addition or loss of a quantity of heat Q(*F).
The above equation can be modified to determine the change in temperature for a fluid stream through a heat exchanger:
- 2) Q = inCpAT Btu where: 6 = Energy addition or rejection rate (Hr )*
Ibm m = Mass flow rate through heat exchanger (gi--).
O AT = remnerature difference of a fiuid stream resuitine frem an energy gain or loss (*F).
Basic equation for boiloff rate determination
- 3) d = in(h -hy) y Btu where: d = Energy addition rate (i.e. Heat Load) (gi:-).
Ibm m = Bolloff rate (Hr ).
(h -h ) = Enthalpy (total energy) g) ease required for vaporization at 212*F. (jncr y f The Heat Exchanger effectiveness of a shell and tube HX w/one shell and even number of tube passes is given below. Ef fectiveness is defined as the ratio of actual heat exchanger heat transfer capability to the maximum theoretical heat transfer possible for given heat exchanger cooling water and hot fluid inlet temperatures.
O
2 1
- 4) c=2 1 + C + (1+C }i 1+exp[-N(1+C){
2 1 - exp [-N(1+CZ)")
where: c = Effectiveness C=e - ratio of hot side (pool water to be cooled) mass flow rate to cooling water flow rate.
N is given by the relationship:
^
N = hcCp where: U = Heat exchanger overall heat transfer coefficient.
This value is referenced by the heat exchanger manufacturer (8tu Hr-Ft
- F),
3 A = Effective heat exchanger heat transfer area. This 2
value is referenced by the heat exchanger manufacturer (Ft ),
once the value of c for the specific heat exchanger Is calculated by equation 4 the various Inlet and outlet heat exchanger temperatures are determined using the following relationships and assumed values:
Thi - Tho a) c=
O Thi - 'ci b) d=mCp(ATc)=mhp(AT)
(equation 2)
C h
c c) Q = mh p(c(Thi - T g))
C e
where: Thi = Hot fluid inlet temperature (assumed fuel pool bulk water temperature) ('F).
T
= Cooling water inlet temperature. This temperature c
is set to the assumed maximum cooling water temperature of 100*F.
Tho = Hot fluid outlet temperature (temperature of fluid being returnec' to pool) (*F).
Since sh p,C and T ; are known, an expression for the equilibrium fuel pool C
c bulkwater temperature (Thg) can be written
- 5) TFuel Pool = Thl = (
)+ 1004 The equations for the effectiveness method of evaluating heat exchanger performance can be found in:
J. P. Holmen, Heat Transfer, 4th ed. (New York: McG raw-H i l l Book Co.,
1976), Chapter 10, Section 6.
l
'TTACHMENT B DUKE POWER COMPANY ARTHUR LOWELL SNOW EDUCATION:
B.S., Nuclear Engineering, University of Tennessee M.E., Mechanical Engineering, University of South Carolina Additional Courses:
Graduate Course Work in Mechanical Engineering toward PhD l
CERTIFICATIONS /PROFES$l0NAL AFFILIATIONS I
Professional Engineer - North Carolina 7397
- South Carolina 6145 Member of the American Nuclear Society YEARS EXPERIENCE:
15
$UMMARY OF PERTINENT EXPERIENCE j
Supervisor Mechanical and Nuclear Division. Nuclear Activities - All Duke Nuclear Power Stations: System-wide radwaste design review activities, licensing activities, probabilistic risk assessment and safety reviews, I
evaluation of nuclear accident scenarlo and corrective actions, radioactive effluents analysis, and nuclear fuel criticality analysis for spent fuel storage designs.
Supervisor. Pipe Support / Restraint Design - Design, engineering and construct-ability of ASME Ill and 831.1.0 piping support / restraints for Catawba Nuclear Station. Development of design criterla and specifications, technical contract administration, scheduling and coordination of design activities.
Assistant Design Engineer - Design and engineering of nuclear fluid process j
systems for Catawba Nuclear Station, review of operating procedures, testing procedures, start up assistance, operating parameters and cost gvaluations.
Assistant Design Enqineer - Development of computer codes for radiation j
shleiding, radioact ve liquid and gaseous discharge.
Preparation of Safety Analysis Reports / Environmental Reports. Radiation shielding designs.
All this work for 0 cones, McGuire and Catawba Nuclear Stations.
EXPERIENCE:
DUKE POWER COMPANY since 1968 1979 to Design Engineer 11 - In charge of Nuclear Sub-Group j
Present responsible for: system-wide redweste design review activities; technical Interface for Mechanical / Nuclear Division with Ilconsing, probabilistic risk assessment and safety review groups; radioactive effluent analysis i
for normal and accident conditions; nuclear fuel crit-l Icelity and generic engineering activities. Generic engineering activities include steam generator chemical 4
cleaning, technical review of responses to TMl concerns, other regulatory and quality assurance matters.
5
. m
__.___.,,---...w.
.-._,,m.#~..--,-.,-,,,-_-,-.-,--.,_-%,
,7-
{}
ARTHUR LOWELL SNOW (Cont'd) 1977-1979 Design Engineer / Assistant Design Engineer - In responsible charge of Pipe Support / Restraint Group for Catawba. Multi-discipline group of Duke and contract personnel included clerks, draftsmen, designers, Mechanical and Civil Engineers (B.S., M.S., and PhD's). Activities included setting up Initial organization, design criteria and specification preparation, contract administration, scheduling, interface with Construction Department, and other Design groups.
Group produced in excess of 30,000 designs for ASME Section 111, Class 2 and 3, and ANSI B31.1.0 piping systems and i
supports for HVAC seismically designed ducting.
l 1972-1977 Assistant Design Engineer - Responsible charge of fluid systems design for all Catawba nuclear process systems.
Supervised preparation of Flow Diagrams, Design Criteria, System Descriptions, Data Sheets for Mechanical Equipment, Safety Analysis Report preparation for Mechanical / Nuclear systems. Developed operating parameters, costs, review of testing / operating procedures, piping system start up as sistan ce.
1968-1972 Assistant Design Engineer / Associate Engineer /Jr. Engineer -
A Responsible charge of Radiation Analysis Group. Activities V
included direct effort and supervisory responsibility for:
development of radiation shleiding computer codes, develop-ment of radioactive liquid and gaseoes discharge computer codes cost evaluation of Turbine-Generator bids using part load heat rates, Safety Analysis Report and Environmental Report preparation, response to NRC (then AEC) questions.
and appearances before ACRS and NRC Staff for Oconee, McGuire and Catawba Nuclear Stations, design radiation shleiding for McGuire and Catawba.
PUBLICATIDNS:
" Criteria For Evaluation of interim Radweste Solidification l
Systens" '83 Weste Management Symposium, Tuscon, Arizona i
2/28/83.
i i
e e
em-w
.-n,
---. - - + -,.--
w --- - -
--n,---w- - -, ---, - -.-,
,we,,.,-ne,,_-,,,--,,ren.,,---.,-c
.e-w-r---------,,,,,-,,e-,,,-we
O i
/
///
e h'/,/
x
'/
/ //
/
C
\\/ / /
=
J i i 5.W(y /\\/ / N W/c s
'y l\\ej/q'.'
e
.C N
/
,g t
W 8
/, O e ~ C,,
/
,' i /
?
I
/
r
/
/
t 0, I l
/
/
l l I!
_Y l
- I'
/
/, )-
j i
~
i l
i (N s'
l jl4 r.
f v Gi ;
i "/
4
/
/
O
- i e1 3
5 i
9 l
-