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NSD-SAE-ESI-97-647 SNC RESPONSE TO NRC RAI ON BELOCA November,1997
NSD-SAE-ESI-97-647 SNC RESPONSE TO NRC RAI ON BELOCA November,1997
                                                        % estinghouse Electric Corpration Energy System Business Unit                                                 ;
% estinghouse Electric Corpration Energy System Business Unit P.O. Ikw 355 Pittsburgh, PA 15230 4 355 01997 Westinghouse Electric Corporation All Rights Reserved 9711250116 971119 ^
P.O. Ikw 355 Pittsburgh, PA 15230 4 355 l
PDR ADOCK 05000348 P
01997 Westinghouse Electric Corporation                                           l All Rights Reserved 9711250116 971119 ^
pg
PDR P          ADOCK 05000348 pg


NSD SAE ESI 97-647
NSD SAE ESI 97-647 12.
: 12.     The plant spectffc modeling and analysis for the large break loss of coolant accident (LOCA) is not provided. Please provide additional information regarding the analysis assumptions for the best estimate large bruk LOCA calcuir .ms. Information needed is the assumed initial conditions, the limiting transient progression with discussion of why it is the limiting transient, singlefailure assumptions, loss-of offsite power assumptions.
The plant spectffc modeling and analysis for the large break loss of coolant accident (LOCA) is not provided. Please provide additional information regarding the analysis assumptions for the best estimate large bruk LOCA calcuir.ms. Information needed is the assumed initial conditions, the limiting transient progression with discussion of why it is the limiting transient, singlefailure assumptions, loss-of offsite power assumptions.
time step assumptions, and major plant parameters with uncenainties. Show that the calculations were performed with the approved version of WCOBRA/ TRAC revision I, andprovide information that shows compliance with the code limitations and restrictions.
time step assumptions, and major plant parameters with uncenainties. Show that the calculations were performed with the approved version of WCOBRA/ TRAC revision I, andprovide information that shows compliance with the code limitations and restrictions.
(WCAP 14723. Section 6 3).
(WCAP 14723. Section 6 3).
Response to RAI-12:
Response to RAI-12:
A summary of results of the Best Estimate (BE) LBLCCA analysis performed for Farley Units and 2 power uprate using ECOBRA/ TRAC was provided in WCAP 14723, Section 6.1.1 (Reference 1) nis analysis followed the approved BE LBLOCA methodology for three and four loop plants (Reference 2). The analysis used ECOBRA/ TRAC MOD 7A, Revision 1, as documented in Reference 3.
A summary of results of the Best Estimate (BE) LBLCCA analysis performed for Farley Units and 2 power uprate using ECOBRA/ TRAC was provided in WCAP 14723, Section 6.1.1 (Reference 1) nis analysis followed the approved BE LBLOCA methodology for three and four loop plants (Reference 2). The analysis used ECOBRA/ TRAC MOD 7A, Revision 1, as documented in Reference 3.
Farley Units I and 2 are three loop plants similar in design to the VRA plant design used to demonstrate the application of the BE LBLOCA methodology in Reference 4. Unit I has an upflow barrel / baffle (B/B) vessel as shown in Figure 12 1 and Unit 2 has a downflow B/B vessel (Figure 12 2). The models are essendally identical except for the presence of gaps 14,15 and 16 which connect the downcomer with the B/B for Unit 2.                 Additionally, the B/B region represented by channel 9 is blocked at the top of Unit 2's section 3. The loop layout for both plants is shown in Figure 12 3. For the purpose of this uprate analysis, it was desirable to perform the full analysis for one unit and have it bound both units. An initial transient was performed for each unit, setting major plant conditions as shown in Table 121. Comparison studies were performed on each plant to determine the limiting configuration. These studies showed that Unit 2 was the limiting plant configuration.
Farley Units I and 2 are three loop plants similar in design to the VRA plant design used to demonstrate the application of the BE LBLOCA methodology in Reference 4. Unit I has an upflow barrel / baffle (B/B) vessel as shown in Figure 12 1 and Unit 2 has a downflow B/B vessel (Figure 12 2). The models are essendally identical except for the presence of gaps 14,15 and 16 which connect the downcomer with the B/B for Unit 2.
With the limiting plant configuration, sensitivity studies were performed to verify the bounding ,
Additionally, the B/B region represented by channel 9 is blocked at the top of Unit 2's section 3. The loop layout for both plants is shown in Figure 12 3. For the purpose of this uprate analysis, it was desirable to perform the full analysis for one unit and have it bound both units. An initial transient was performed for each unit, setting major plant conditions as shown in Table 121. Comparison studies were performed on each plant to determine the limiting configuration. These studies showed that Unit 2 was the limiting plant configuration.
With the limiting plant configuration, sensitivity studies were performed to verify the bounding,
plant conditions. A Loss-Of-Offsite Power (LOOP) study confirmed the assumption that pumps offsite power available is more limiting. A SI injection study showed that the assumption of loss of a full train of SI was the limiting single failure assumption. These assumptions were used in the reference transient.
plant conditions. A Loss-Of-Offsite Power (LOOP) study confirmed the assumption that pumps offsite power available is more limiting. A SI injection study showed that the assumption of loss of a full train of SI was the limiting single failure assumption. These assumptions were used in the reference transient.
Initial calculations were performed using a CD=1.0 DECLG break. Further calculations showed inat a CD=1.0 Split break was the limiting reference transient. The results of these calculations are shown in Table 12-2. A plot of the PCT transient for the CD=1.0 Split case is shown in Figure 12-4. Split breaks have been determined to be more limiting in some cases because they result in a small downward core flow during blowdown (see following discussion).
Initial calculations were performed using a CD=1.0 DECLG break. Further calculations showed inat a CD=1.0 Split break was the limiting reference transient. The results of these calculations are shown in Table 12-2. A plot of the PCT transient for the CD=1.0 Split case is shown in Figure 12-4. Split breaks have been determined to be more limiting in some cases because they result in a small downward core flow during blowdown (see following discussion).
Line 38: Line 40:


NSD-SAE ESI-97-647
NSD-SAE ESI-97-647
              ' Reference Transient Description The LOCA transient can be divided into time periods in which specific phenomena are occurring.
' Reference Transient Description The LOCA transient can be divided into time periods in which specific phenomena are occurring.
            - A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the hot assembly undergoes. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below. Results are shown in Figures 12-4 to 1217.
- A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the hot assembly undergoes. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below. Results are shown in Figures 12-4 to 1217.
Critical Heat Flux (CHF) Phase 4
Critical Heat Flux (CHF) Phase Immediately following the cold leg rupture, the break flowrate is'subcoc'ed and high. The 4
Immediately following the cold leg rupture, the break flowrate is'subcoc'ed and high. The regions of the RCS with the hottest initial temperatures (core, upper plenum, upper head, and hot legs) begin to flash to steam within the first 0.5 seconds following the break. Flow in the core reverses, and the fuel rods begin to go through departure from nucleate boiling (DNB). Voiding in the core also causes the fission power to drop rapidly. The discharge flowrate decreases sharply as the break flow becomes two-phase (Figure 12 6). This phase is terminated when the water in the lower plenum and downcomer (DC) begin to fla)h.
regions of the RCS with the hottest initial temperatures (core, upper plenum, upper head, and hot legs) begin to flash to steam within the first 0.5 seconds following the break. Flow in the core reverses, and the fuel rods begin to go through departure from nucleate boiling (DNB). Voiding in the core also causes the fission power to drop rapidly. The discharge flowrate decreases sharply as the break flow becomes two-phase (Figure 12 6). This phase is terminated when the water in the lower plenum and downcomer (DC) begin to fla)h.
Uoward Core Flow Phase Flashing in the lower plenum and pumped flow supplied by the intact loops re establishes upward core flow for a brief period of time (Figure 12 7). This phase ends as the lower plenum mass is depleted, the loops become two phase, and the intact loop pump head degrades because of two-phase conditions (Figure 12 8).
Uoward Core Flow Phase Flashing in the lower plenum and pumped flow supplied by the intact loops re establishes upward core flow for a brief period of time (Figure 12 7). This phase ends as the lower plenum mass is depleted, the loops become two phase, and the intact loop pump head degrades because of two-phase conditions (Figure 12 8).
Downward Core Flow Phase Downward flow into the core begins as the pump head continues to be degraded and upward flow in the DC is firmly established (Figure 12-9).
Downward Core Flow Phase Downward flow into the core begins as the pump head continues to be degraded and upward flow in the DC is firmly established (Figure 12-9).
Due to the downflow during this phase, tne cladding temperature was turned around at about 15 secunds after the initiation of the transient. For certain size split breaks, the fraction of break flow drawn from the core is smaller than the DECLG break, leading to poorer core cooling during this period. The accumulators on the intact loops begin to inject at 14 seconds after the break (Figure 12-10). Initially, the injected water is bypassed around the downcomer and out of the break. As the system pressure continues to' fall (Figure 12-11), the break flow and consequently the downward core flow are reduced. The vessel pressure reaches the containment
Due to the downflow during this phase, tne cladding temperature was turned around at about 15 secunds after the initiation of the transient. For certain size split breaks, the fraction of break flow drawn from the core is smaller than the DECLG break, leading to poorer core cooling during this period. The accumulators on the intact loops begin to inject at 14 seconds after the break (Figure 12-10). Initially, the injected water is bypassed around the downcomer and out of the break. As the system pressure continues to' fall (Figure 12-11), the break flow and consequently the downward core flow are reduced. The vessel pressure reaches the containment pressure at the end of this phase, which occurs about 30 seconds after the initiation of the transient. The core begins to heat up as the system approaches containment pressure and the vessel begins to fill with ECCS water.
,        pressure at the end of this phase, which occurs about 30 seconds after the initiation of the transient. The core begins to heat up as the system approaches containment pressure and the vessel begins to fill with ECCS water.
Refill Phase When the steam flow up the downcomer is sufficiently : iced, itc cold ECCS water begins to penetrate the downcomer (Figure 12-12) and refill the lower plenum (Figure 12-13). The refill 12 2
Refill Phase When the steam flow up the downcomer is sufficiently : iced, itc cold ECCS water begins to penetrate the downcomer (Figure 12-12) and refill the lower plenum (Figure 12-13). The refill 12 2
  ..,mw -    .      .,4.s. a.
..,mw
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a.
A
A


NSD SAE ESI 97-647 design, which is covered by the SER.. The temperature ranges used for the accumulator and SI ECCS water were calculated using plant specific data collected for a year to determine the nominal and maximum values for ranging. All requirements outlined in the SER were confirmed for the Farley analysis. Among the required checks are the following:
NSD SAE ESI 97-647 design, which is covered by the SER.. The temperature ranges used for the accumulator and SI ECCS water were calculated using plant specific data collected for a year to determine the nominal and maximum values for ranging. All requirements outlined in the SER were confirmed for the Farley analysis. Among the required checks are the following:
: 1. Confirm all transient runs predict cladding burst when PCT exceeds 1600'F.
1.
: 2. Confirm normality of several key distributions used in the analysis.
Confirm all transient runs predict cladding burst when PCT exceeds 1600'F.
The required calculation of core wide oxidation has been performed and the result is shown on Table 6.1.12 of Reference 1.           The long term core cooling calculation did not use ECOBRAfrRAC and is discussed further in the response to RAI 18.
2.
Confirm normality of several key distributions used in the analysis.
The required calculation of core wide oxidation has been performed and the result is shown on Table 6.1.12 of Reference 1.
The long term core cooling calculation did not use ECOBRAfrRAC and is discussed further in the response to RAI 18.
The time step assumptions used in the Parley analysis are listed in Table 12 3 and follow the approved methodology.
The time step assumptions used in the Parley analysis are listed in Table 12 3 and follow the approved methodology.
Summary The BE LBLOCA analysis for Farley Units 1 and 2 followed all of the guidelines and requirements as specified in the SER (Reference 2) and resulted in a peak cladding temperature
Summary The BE LBLOCA analysis for Farley Units 1 and 2 followed all of the guidelines and requirements as specified in the SER (Reference 2) and resulted in a peak cladding temperature
  <2064'F, which meets the acceptance criteria. The operating ranges for major plant parameters are accounted for in the estimated PCT uncertainty. Table 12-2 shows some of these parameter ranges. A complete list is provided in the FSAR and in the response to RAI 13.
<2064'F, which meets the acceptance criteria. The operating ranges for major plant parameters are accounted for in the estimated PCT uncertainty. Table 12-2 shows some of these parameter ranges. A complete list is provided in the FSAR and in the response to RAI 13.
12-4 s
12-4 s


                                                                              - NSD-SAE ESI 97-647
- NSD-SAE ESI 97-647


==References:==
==References:==
: 1. WCAP-14723, "Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report," January,1997,
1.
: 2.     Letter, R. C. Jones (USNRC) to N. J. Liparulo (E), " Acceptance for Referencing of the Topical Report WCAP 12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of Coolant Analysis," June 28,1996.
WCAP-14723, "Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report," January,1997, 2.
: 3. Letter, N. J. Liparuto (_W) to R. C. Jones (USNRC), " Revisions to Westinghouse Best.
Letter, R. C. Jones (USNRC) to N. J. Liparulo (E), " Acceptance for Referencing of the Topical Report WCAP 12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of Coolant Analysis," June 28,1996.
3.
Letter, N. J. Liparuto (_W) to R. C. Jones (USNRC), " Revisions to Westinghouse Best.
Estimate Methodology," NTD-NRC 95 4575, October 13, 1995.
Estimate Methodology," NTD-NRC 95 4575, October 13, 1995.
: 4.     " Westinghouse Code Qualification Document for Best Estimate Loss of coolant Accident Analysis," WCAP-12945 P (Proprietary), Volumes I-V.
4.
" Westinghouse Code Qualification Document for Best Estimate Loss of coolant Accident Analysis," WCAP-12945 P (Proprietary), Volumes I-V.
12-5
12-5


A NSD-SAE ESI-97-647 '
A NSD-SAE ESI-97-647 '
Lt                                     Table 121 -
Lt Table 121 -
                                                  . Major Plant Parameter Initial Assumptions Used in the BE LBLOCA Analysis-                                                                                                       !
. Major Plant Parameter Initial Assumptions Used in the BE LBLOCA Analysis-for Farley Units I and 2 Power Uprate;
for Farley Units I and 2 Power Uprate;
^
                                                                                                                                                                                                                        ^
Parameter''
Parameter''                                             Initial Value Range Plant Initial Operating Conditions:                                                                                                                                                     I Reactor Power                                                                   100% of                                     5102% of
Initial Value Range Plant Initial Operating Conditions:
                                                                                                                                  .2775 MWt                                         -2775 MWt -
I Reactor Power 100% of 5102% of
l Peak Linear Heat Rate (PLHR)                                     Derived from desired                                           Fo's 2.5 Tech Spec (TS)!!mit
.2775 MWt
;-                                                                                                                              Fn= 2.5 and maximum baseload Fo Hot Rod Avg Linear Heat Rate                                       Derived from TS                                         Fas 1.70 F ,= 1.7 Fluid Conditions Tavg                                                                           567.2*F                         567.2 2 6'F s Tavg 5 577.2
-2775 MWt -
* 6*F Pressurizer Pressure                                                       2250 psia                                   2250
l' Peak Linear Heat Rate (PLHR)
Derived from desired Fo's 2.5 Tech Spec (TS)!!mit F = 2.5 and n
maximum baseload Fo Hot Rod Avg Linear Heat Rate Derived from TS Fas 1.70 F,= 1.7 Fluid Conditions Tavg 567.2*F 567.2 2 6'F s Tavg 5 577.2
* 6*F Pressurizer Pressure 2250 psia 2250
* 50 psi
* 50 psi
                                                      - Loop Flow -                                                         86,000 GPM                                       2: 86,000 GPM                                         '
- Loop Flow -
Accumulator Temperature                                                             105'F                           90s T.s 120*F
86,000 GPM 2: 86,000 GPM Accumulator Temperature 105'F 90s T.s 120*F Accumulator Pressure 640 psia 600 s P, s 680 psia Accumulator Volume 980 ft' 965 s V.5995 ft' Accident Boundary Conditions s
,                                                      Accumulator Pressure                                                         640 psia                         600 s P, s 680 psia Accumulator Volume                                                               980 ft'                           965 s V.5995 ft' s                                        Accident Boundary Conditions j                                                     Offsite Power Availability                                                               On                                   On or Off Single Failure Assumption                                           Loss of one RHR                                                   -
j Offsite Power Availability On On or Off Single Failure Assumption Loss of one RHR pump Safety Injection Temperature Nominal (85'F)
pump Safety Injection Temperature                                         Nominal (85'F)                       - 70 s SI Temp s 100 'F Safety Injection Delay .                                                         12 sec.                             512 seconds (no LOOP) s 27 sec. (LOOP)
- 70 s SI Temp s 100 'F Safety Injection Delay.
Containment Pressure -                                                     - Bounded                                   See Figure 12-17 i
12 sec.
Steam Generator Tube Plugging                                                         20%                                         0-20 %
512 seconds (no LOOP) s 27 sec. (LOOP)
Containment Pressure -
- Bounded See Figure 12-17 i
Steam Generator Tube Plugging 20%
0-20 %
+
+
12-6 s -fr--e   '-- -
12-6 s
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w


NSD-SAE-ESI 97-647 Table 12 2 Scoping Study Results for Farley Unit 2 Break Type             Break Size'       Renood 1 PCT ('F)     Renood 2 PCT
NSD-SAE-ESI 97-647 Table 12 2 Scoping Study Results for Farley Unit 2 Break Type Break Size' Renood 1 PCT ('F)
Renood 2 PCT
('F)
('F)
DECLG                     1,0               1810                 1861 Split                   0.8                 1718                 1650 Split                   1.0                 1871                 1936 Split                   1.2                 1867                 1778 Split                   1.4                 1718               1436
DECLG 1,0 1810 1861 Split 0.8 1718 1650 Split 1.0 1871 1936 Split 1.2 1867 1778 Split 1.4 1718 1436
* Fraction of cold leg area (CD for DECLG)
* Fraction of cold leg area (CD for DECLG)
Table 12 3 Time Step Assumptions for Farley Unit 2 Calculations
Table 12 3 Time Step Assumptions for Farley Unit 2 Calculations 8,C 12 7
-                                                                                        8,C 12 7


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NSD-SAE.ESI.97-647
NSD-SAE.ESI.97-647
        .f. 3 4
.f. 3 4
5                                                                                           s.e.,on (2s n E .r                                                     2                                           -
-5 s.e.,on.
s s          -v. E.
(2s n E
s.e.on
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                                                                                                                                                                                                                <= m e, n -                ::, . am. L                                                                                             _......                                                              .
-v. E s.e.on
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4
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j
                                                                          -0' sa.                                                                                   ;:
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-3 3
                                                    , .s. . ..,_...O,.......,.......................
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    ,p,.                 .. . . . . '                          . . n . .* . g . . . .. .e .. .- .. . . . . . . . . . . . . . . . . . . . . . . . .                                  .    .M.. ..            secean 3 5 310 ==
.......,,...............,....4....+...
LP                           MA : om                                                   ! Lp 5..          .**
3,,an g a
* 0 *'' M
+
    , , , , , _ .      o W=......."#...=    .      .
"..........-4y am Si E !E! S...==:=.i E E ; E! E
                                                                .*.*. .."*'..N. **4+e***"*..*.***--*.+gy.
: E :E! E dif4if I
                                                                                                                                                .e *- .Q.a .* .*..-...Q.....".
... a:s..... m...i.... m e ::=::...#-.D' O. g.. = :.h. 3
                                                                                                                                                                          . . : .'. f.../.*..G.
+E...;. 3.Y. 9 s_
g.. . .. . . . . . . .O. . .. . . .
r:6% 5.. ".. _.
    ,u      _.,          .......                     ..     ......+.........,..........+...............                                                                  . . . . .....
i
                          ........                    ........................................t..................;...........
: 4. 9-3'..... :..... ;.....:.......... :.......... ;........-4 t....
* A p
................. +............................
WW Gf . &su um aus em nas as as su um um sus em a 4                                                                               0 G3          .
..........,s.....;.......
g Secaon t mm
m,.
                                                          . . .            . . . . . .              . . . . . .          . . . . . .        . . . . .              . . j. . . '    . . . . . .
,,...,_...O,.......,.......................
                      #. 4- 8                     .
,p,.
4     .
,.s..
G                                                 Q cr. anne                               s.c on i
.M....
                                                                                                                                                                                                              <. m a                                                                                                                                                            O cm                                         ,
secean 3
                                                ~ ..
.. n... g......e...-..
LP MA : om
! Lp 0 *'' M
.*.*." N *4+e***"*..*.***--*.+gy..Q..... :.....G...........O.......  
.e *- a * *..-...Q.....".
.' f. /.* g..
5 310==
o W=......."#...=.. *..
* 5..
...... +.........,.......... +...............
,u A
........................................t..................;...........
p Gf. &su um aus em nas as as su um um sus em a WW 4
G3 0
g Secaon t j
mm 4-8 4
G Q cr. anne s.c on i
<. m O cm a
~..
Figure 12 2. Farley Unit 2.W_ COBRA / TRAC Vevel Noding (Vertical View) 12 9
Figure 12 2. Farley Unit 2.W_ COBRA / TRAC Vevel Noding (Vertical View) 12 9


NSD-SAE ESt-97-647 LOOP 2           ru, sas*
NSD-SAE ESt-97-647 LOOP 2 ru, sas*
g~                         n" canomen puw                                                  O Junaa
g~
                                        ,          3 E
n canomen O Junaa puw 3
f 4                     4 3       eness.
E f
j                         E u:cun s,                               Twout meuu si n.L     13I                                y ES' 9                       8 g
4 4 3 eness.
yN
j E
                                        ,    VESSEL 7     g i
u:cun s, Twout meuu 13I y ES' si n.L 9
                                                                        + M'                 g puw r e sLE               . g             _
8g yN puw VESSEL i
i v                '
7 g
( p SREM -
+ M' g
E                     pg4x resLE O             ''
r e sLE
n.L pgT$'       N                                                               i as                                                   d       pgAK 7
. g i
LOOP 3                ~
(
Oe PUW                                          O ;*a**
p v
Figure 12 3. Farley Units 1 and 2 ECOBRA/ TRAC Loop Layout (Steady State) 12 10
E pg4x O
SREM -
resLE n.L pg $'
N T
i as d pgAK LOOP 3 7
Oe
~
O
;*a**
PUW Figure 12 3. Farley Units 1 and 2 ECOBRA/ TRAC Loop Layout (Steady State) 12 10


l.
l.
NSD-SAE ESI 97-647 2000                                                                                 11.5 e_   ,
NSD-SAE ESI 97-647 2000 11.5 e_
1 M                                           :
1 M
                                                                                              - 11
]f
                                    )
- 11 I
                                                                              ]f I
)
p1
p1
[                                           iL             .
[
1500       ,
iL 1500
__[
__[
_3o,3 l                                         .
_3o,3 l
n                                                                                     -
l n
m l                                                  -
m 1
1
- 10 I
* I                                          - 10 e         -
e i
          = 1000                                  i                                          .        =
=
a                                   ,-          '
=
i                                                         =
1000 a
c                                  i
i
                                                                                              .,,s s         .                        s                                               -
=
o e                                 i                                                         -
i
e                                                                                   .
.,,s c
s          -                        i                                                       w i                                                -
s s
I                                                -
o e
n I                        '                                                .s 500 . _s                __a 1          l l          1                                                      -
i e
i,- -                                                          - e,s
i w
                              ,I1                                                            -
s i
s i
                ,         ,iii         iiii       ,,ii             iiii   .iii       iiil    a 9               50           100         100         200       2U0     300 TlWE (seconds)
I n
.s I
500. _s
__a 1
l l
1 i,- -
- e,s
,I1 i
s iiil
,iii iiii
,,ii iiii
.iii a
9 50 100 100 200 2U0 300 TlWE (seconds)
Figure 12 4. PCT for Farley Reference Transient, CD=1.0 Cold Leg Split Break 12 11
Figure 12 4. PCT for Farley Reference Transient, CD=1.0 Cold Leg Split Break 12 11


NSD-SAE ESI-97-647
NSD-SAE ESI-97-647 HR TCLAD et 5 92 ft.
          -~~~HR HR TCLAD et 5 92 ft.
-~~~HR TCLAD et 8.08 ft.
TCLAD et 8.08 ft.
~~~~HR TCLAD et 10 33 ft.
          ~~~~HR         TCLAD et 10 33 ft.
2000 s
2000 rs    ,
r Y'%,.
Y'%,.               ,
"p n*.
                                                                      ''            "p n*.
lw}4
                    .      pt    lw}4           %                                            '\ s 1500 A #   '
'\\
                                .r,r                 ',                                              ''
t p
ylo- h t                              's                                                 >
s A #
      ^                                                   's v
1500 ylo-
      ~                                                       \
.r,r h
taag, s
's t
e             .                                                        %
's
w                                                                       ,
^
3 1000                                                                     \
v
o                                                                             *g u           -
\\
~
: taag, s
e w
3 1000
\\
o
*g u
s e
s e
is\\
cm.
cm.
E is\
t E
t 1
1 e
e          a                                                                           g H                                                                                       I
a g
* i 94 500                       - .                                                        '
H I
                  -                          i W
i 94 500 W
g 1
i g
                                                                  ' ' -                          Ltts- ,
1 Ltts-,
0 O           50           Illo             100             200               200             3W0 TlWE (seconds)
0 O
50 Illo 100 200 200 3W0 TlWE (seconds)
Figure 12 5. Cladding Temperature at Selected Elevations for CD=1.0 Split 12 12
Figure 12 5. Cladding Temperature at Selected Elevations for CD=1.0 Split 12 12


NSD-SAE ESI 97 647 50000-40000 o             ,.
NSD-SAE ESI 97 647 50000-40000 o
N E 30000                               -
NE 30000 4
4             .
v o
v             .
w o
o           .
20000 s
w o   20000 s             -
o 6
o 6
m   10000 m
m 10000 m
o           -
o
:M           -
:M
(
~
              ~
t (
t       .1 , ,a                       _
.1,
    -10000       ''''      ''''      ''''          ''''    ''''      ''''
,a
O            50       tUO         100       200       200       300 TIME (seconds)
-10000 O
50 tUO 100 200 200 300 TIME (seconds)
Figure 12 6 Break Flow for CD=1.0 Split 12 13
Figure 12 6 Break Flow for CD=1.0 Split 12 13


NSD-SAE ESI 97-647 20000-Im 10000
NSD-SAE ESI 97-647 20000-Im 10000
        ^                                                                           ;
^
o             .
o u
u n
n N
N               '
E a
E a               .
w e
w e
        ~         0                  ,      -
0
6             "
~
m o
6 mo s
s
~
                        ~
n n
n n
a             -
a 2
2
-10000 7
            -10000         7
~
                      ~
-20000 O
            -20000           ' ' ' '    ''''        ' ' ' '      ' ' ' '
O 10 15 20 TlWE (seconds)
O               O          10         15               20 TlWE (seconds)
Figure 12 7. Flow at the Bottom of tue Core for CD=1.0 Split 12 14
Figure 12 7. Flow at the Bottom of tue Core for CD=1.0 Split 12 14


NSD-SAE-ESI 97-647 1
NSD-SAE-ESI 97-647 1
f
f
        .8 e
.8 e
o .6 u
o.6 u
O m         .
O m
6 o .+
6
.o.+
2
2
              ' '  '''    ''''    ''''        ''''    ''''    ''i' 0
''i' 0
O         50     140         !!io     208     200       300 TlWE (seconds)
O 50 140
!!io 208 200 300 TlWE (seconds)
Figure 12 8. Intact Reactor Coolant Pump Void Fraction for CD=1.0 Split 12 15
Figure 12 8. Intact Reactor Coolant Pump Void Fraction for CD=1.0 Split 12 15


NSD-SAE ESI 97-647 8000 ..
NSD-SAE ESI 97-647 8000..
4000 m
4000 m
u             -
u o
o M   2000                                                                                       '
M 2000 N
N E             -
E
          .o v
.o v
S           a
S
                                      ^
^
o                                 .
a o
e o            -
eo n -2000 M
n -2000 M
o 2
o           .
-4000
2
-sooO O
                -4000
O 10 IS 20 TIME (seconds)
                -sooO         ' ' ' '    ''''            ' ' ' '        ' ' ' '
O               O            10           IS                     20 TIME (seconds)
Figure 12 9 Blowdown Flow at the Top of the Core for CD=1.0 Split 12 16 g;
Figure 12 9 Blowdown Flow at the Top of the Core for CD=1.0 Split 12 16 g;
                                                                      ..n                                 -.
..n


  .~ - .. . - - _ - .          ~ _ .          . . .    -    .
.~
                                                                .  .. - - . .        - -. - .... . - ~ .                  - . -      ..
~ _.
- -. -..... - ~.
P NSD-SAE-ESI 97-647
P NSD-SAE-ESI 97-647
                          - latest Assemuletgr
- latest Assemuletgr
                          - - - - Bretse Assemulater 3000
- - - - Bretse Assemulater 3000 2000 m
* 2000                                                                                                             !
M N
m               .
M N               "
E 2000 -
E 2000 -
                      .a
.a
                      ~             .
~
v             .
v m
m a 1500 as-           ,
a 1500 as-m o
m           -
6 m 1000 p
o             -
:s 500 i
6             .
ie i f\\iei g
m 1000 p
ii - ie i i i i i 1.
:s           -
>>,i Li 50 lite 100 200 200 36l0 TlWE (seconds)
500 g        i   ie i     f\iei       ii - ie i i       i i i       1.         >>,i Li                   50     lite         100   200               200               36l0 TlWE (seconds)
Figure 1210. Accumulator Flow for CDal.0 Sput i
Figure 1210. Accumulator Flow for CDal.0 Sput i
12-17
12-17
Line 440: Line 552:
NSD SAE ESI 97 647 2500 2000 m
NSD SAE ESI 97 647 2500 2000 m
o
o
  *~
*~
m 1500 Q-         ..
m 1500 Q-v o
v o         _                          ,
=
  =         _
M 1000 o.,
M
500 I
  " 1000 o.,
I I
500
I I
                  ! ! $    I I I I     I   I I I     I  u I     x x x I     r I I !
I I
ll           80         100           !!l0       200         2!ie           3U0 T!WE (seconds)
I I
u I
x x
x I
r I
I ll 80 100
!!l0 200 2!ie 3U0 T!WE (seconds)
Figure 1211. Pressurizer Pressure for CD=1.0 Split 12 18
Figure 1211. Pressurizer Pressure for CD=1.0 Split 12 18


NSD-SAE ESI-97-647 30 L
NSD-SAE ESI-97-647 30 L
25
25
                          ~
~
m         ~
m
C         _
~
_      l            l 2,            I       ll11.111
C l
                        ~
l I
                                                                .. o l!'b                 Mk l              E''     _
ll11.11 1
lj Z         _
2,
0         -
~
                    '        F t
o l!'b Mk lj E''
o
l Z
                            ''''      ''ii           iiii       iie i       ,,ii       iiii u           is           ido         Ino         290     2tio           suo TlWE (seconds)
0 F
t
''ii iiii iie i
,,ii iiii o
u is ido Ino 290 2tio suo TlWE (seconds)
Figure 1212. Average Collapsed Liquid Level in the Downcomer for CD=1.0 Spilt 12 19
Figure 1212. Average Collapsed Liquid Level in the Downcomer for CD=1.0 Spilt 12 19


NSD-SAE ESI 97-647 10                                         __
NSD-SAE ESI 97-647 10 T j 'ig
T
%.~
                              ,        j 'ig                                               %.~
e m
e m
          ~
~
W o
W o
e           8 a
8 e
v c-J w             '
a v
        .o in               -
c-J w
m.
.o in m.
o                 .
o o
o                 -
o 2
o 2
g          iie i     iei     eii i         e i i i           iii,     ei ii LI           50     1l10         llie         29             2li0         300 TlWE (seconds Figure 12-13. Couapsed Liquid Level in the Lower Plenum for CD=1.0 Split l
iie i iei eii i e i i i
: iii, ei ii g
LI 50 1l10 llie 29 2li0 300 TlWE (seconds Figure 12-13. Couapsed Liquid Level in the Lower Plenum for CD=1.0 Split l
12 20 l
12 20 l


NSD-SAE ESI 97-647 180000 -
NSD-SAE ESI 97-647 180000 -
180000 t40000                                                                 -
180000 t40000 120000
120000
^
        ^                 .
E
E
        .o                 -
.o C
C               '
100000 l
100000               l                                                   -
/
B0000                                       y
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Figure 1214. Vessel Water Mass for CD=1.0 Split 12 21
Figure 1214. Vessel Water Mass for CD=1.0 Split 12 21


NSD-SAE-ESI 97-647 Lever Power Chesnel 10
NSD-SAE-ESI 97-647 Lever Power Chesnel 10
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Figure 1216. Pumped Safety Injection thw for CD=1.0 Split 12 23 i


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F Figure 1217. Containment Backpressure for CD=1.0 Split 12 24
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Latest revision as of 07:36, 10 December 2024

Non-proprietary NSD-SAE-ESI-97-647 to SNC Response to NRC RAI on Beloca
ML20199G187
Person / Time
Site: Farley  Southern Nuclear icon.png
Issue date: 11/19/1997
From:
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19313D093 List:
References
NSD-SAE-ESI-97, NSD-SAE-ESI-97-647, NUDOCS 9711250116
Download: ML20199G187 (24)
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Text

-

- ~ ~ - - - "

- ~ ~ ' ~ ~ " ^

^

NSD-SAE-ESI-97-647 SNC RESPONSE TO NRC RAI ON BELOCA November,1997

% estinghouse Electric Corpration Energy System Business Unit P.O. Ikw 355 Pittsburgh, PA 15230 4 355 01997 Westinghouse Electric Corporation All Rights Reserved 9711250116 971119 ^

PDR ADOCK 05000348 P

pg

NSD SAE ESI 97-647 12.

The plant spectffc modeling and analysis for the large break loss of coolant accident (LOCA) is not provided. Please provide additional information regarding the analysis assumptions for the best estimate large bruk LOCA calcuir.ms. Information needed is the assumed initial conditions, the limiting transient progression with discussion of why it is the limiting transient, singlefailure assumptions, loss-of offsite power assumptions.

time step assumptions, and major plant parameters with uncenainties. Show that the calculations were performed with the approved version of WCOBRA/ TRAC revision I, andprovide information that shows compliance with the code limitations and restrictions.

(WCAP 14723. Section 6 3).

Response to RAI-12:

A summary of results of the Best Estimate (BE) LBLCCA analysis performed for Farley Units and 2 power uprate using ECOBRA/ TRAC was provided in WCAP 14723, Section 6.1.1 (Reference 1) nis analysis followed the approved BE LBLOCA methodology for three and four loop plants (Reference 2). The analysis used ECOBRA/ TRAC MOD 7A, Revision 1, as documented in Reference 3.

Farley Units I and 2 are three loop plants similar in design to the VRA plant design used to demonstrate the application of the BE LBLOCA methodology in Reference 4. Unit I has an upflow barrel / baffle (B/B) vessel as shown in Figure 12 1 and Unit 2 has a downflow B/B vessel (Figure 12 2). The models are essendally identical except for the presence of gaps 14,15 and 16 which connect the downcomer with the B/B for Unit 2.

Additionally, the B/B region represented by channel 9 is blocked at the top of Unit 2's section 3. The loop layout for both plants is shown in Figure 12 3. For the purpose of this uprate analysis, it was desirable to perform the full analysis for one unit and have it bound both units. An initial transient was performed for each unit, setting major plant conditions as shown in Table 121. Comparison studies were performed on each plant to determine the limiting configuration. These studies showed that Unit 2 was the limiting plant configuration.

With the limiting plant configuration, sensitivity studies were performed to verify the bounding,

plant conditions. A Loss-Of-Offsite Power (LOOP) study confirmed the assumption that pumps offsite power available is more limiting. A SI injection study showed that the assumption of loss of a full train of SI was the limiting single failure assumption. These assumptions were used in the reference transient.

Initial calculations were performed using a CD=1.0 DECLG break. Further calculations showed inat a CD=1.0 Split break was the limiting reference transient. The results of these calculations are shown in Table 12-2. A plot of the PCT transient for the CD=1.0 Split case is shown in Figure 12-4. Split breaks have been determined to be more limiting in some cases because they result in a small downward core flow during blowdown (see following discussion).

12 1

NSD-SAE ESI-97-647

' Reference Transient Description The LOCA transient can be divided into time periods in which specific phenomena are occurring.

- A convenient way to divide the transient is in terms of the various heatup and cooldown phases that the hot assembly undergoes. For each of these phases, specific phenomena and heat transfer regimes are important, as discussed below. Results are shown in Figures 12-4 to 1217.

Critical Heat Flux (CHF) Phase Immediately following the cold leg rupture, the break flowrate is'subcoc'ed and high. The 4

regions of the RCS with the hottest initial temperatures (core, upper plenum, upper head, and hot legs) begin to flash to steam within the first 0.5 seconds following the break. Flow in the core reverses, and the fuel rods begin to go through departure from nucleate boiling (DNB). Voiding in the core also causes the fission power to drop rapidly. The discharge flowrate decreases sharply as the break flow becomes two-phase (Figure 12 6). This phase is terminated when the water in the lower plenum and downcomer (DC) begin to fla)h.

Uoward Core Flow Phase Flashing in the lower plenum and pumped flow supplied by the intact loops re establishes upward core flow for a brief period of time (Figure 12 7). This phase ends as the lower plenum mass is depleted, the loops become two phase, and the intact loop pump head degrades because of two-phase conditions (Figure 12 8).

Downward Core Flow Phase Downward flow into the core begins as the pump head continues to be degraded and upward flow in the DC is firmly established (Figure 12-9).

Due to the downflow during this phase, tne cladding temperature was turned around at about 15 secunds after the initiation of the transient. For certain size split breaks, the fraction of break flow drawn from the core is smaller than the DECLG break, leading to poorer core cooling during this period. The accumulators on the intact loops begin to inject at 14 seconds after the break (Figure 12-10). Initially, the injected water is bypassed around the downcomer and out of the break. As the system pressure continues to' fall (Figure 12-11), the break flow and consequently the downward core flow are reduced. The vessel pressure reaches the containment pressure at the end of this phase, which occurs about 30 seconds after the initiation of the transient. The core begins to heat up as the system approaches containment pressure and the vessel begins to fill with ECCS water.

Refill Phase When the steam flow up the downcomer is sufficiently : iced, itc cold ECCS water begins to penetrate the downcomer (Figure 12-12) and refill the lower plenum (Figure 12-13). The refill 12 2

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A

NSD SAE ESI 97-647 design, which is covered by the SER.. The temperature ranges used for the accumulator and SI ECCS water were calculated using plant specific data collected for a year to determine the nominal and maximum values for ranging. All requirements outlined in the SER were confirmed for the Farley analysis. Among the required checks are the following:

1.

Confirm all transient runs predict cladding burst when PCT exceeds 1600'F.

2.

Confirm normality of several key distributions used in the analysis.

The required calculation of core wide oxidation has been performed and the result is shown on Table 6.1.12 of Reference 1.

The long term core cooling calculation did not use ECOBRAfrRAC and is discussed further in the response to RAI 18.

The time step assumptions used in the Parley analysis are listed in Table 12 3 and follow the approved methodology.

Summary The BE LBLOCA analysis for Farley Units 1 and 2 followed all of the guidelines and requirements as specified in the SER (Reference 2) and resulted in a peak cladding temperature

<2064'F, which meets the acceptance criteria. The operating ranges for major plant parameters are accounted for in the estimated PCT uncertainty. Table 12-2 shows some of these parameter ranges. A complete list is provided in the FSAR and in the response to RAI 13.

12-4 s

- NSD-SAE ESI 97-647

References:

1.

WCAP-14723, "Farley Nuclear Plant Units I and 2 Power Uprate Project NSSS Licensing Report," January,1997, 2.

Letter, R. C. Jones (USNRC) to N. J. Liparulo (E), " Acceptance for Referencing of the Topical Report WCAP 12945 (P), Westinghouse Code Qualification Document for Best Estimate Loss-of Coolant Analysis," June 28,1996.

3.

Letter, N. J. Liparuto (_W) to R. C. Jones (USNRC), " Revisions to Westinghouse Best.

Estimate Methodology," NTD-NRC 95 4575, October 13, 1995.

4.

" Westinghouse Code Qualification Document for Best Estimate Loss of coolant Accident Analysis," WCAP-12945 P (Proprietary), Volumes I-V.

12-5

A NSD-SAE ESI-97-647 '

Lt Table 121 -

. Major Plant Parameter Initial Assumptions Used in the BE LBLOCA Analysis-for Farley Units I and 2 Power Uprate;

^

Parameter

Initial Value Range Plant Initial Operating Conditions:

I Reactor Power 100% of 5102% of

.2775 MWt

-2775 MWt -

l' Peak Linear Heat Rate (PLHR)

Derived from desired Fo's 2.5 Tech Spec (TS)!!mit F = 2.5 and n

maximum baseload Fo Hot Rod Avg Linear Heat Rate Derived from TS Fas 1.70 F,= 1.7 Fluid Conditions Tavg 567.2*F 567.2 2 6'F s Tavg 5 577.2

  • 6*F Pressurizer Pressure 2250 psia 2250
  • 50 psi

- Loop Flow -

86,000 GPM 2: 86,000 GPM Accumulator Temperature 105'F 90s T.s 120*F Accumulator Pressure 640 psia 600 s P, s 680 psia Accumulator Volume 980 ft' 965 s V.5995 ft' Accident Boundary Conditions s

j Offsite Power Availability On On or Off Single Failure Assumption Loss of one RHR pump Safety Injection Temperature Nominal (85'F)

- 70 s SI Temp s 100 'F Safety Injection Delay.

12 sec. 512 seconds (no LOOP) s 27 sec. (LOOP)

Containment Pressure -

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Renood 2 PCT

('F)

DECLG 1,0 1810 1861 Split 0.8 1718 1650 Split 1.0 1871 1936 Split 1.2 1867 1778 Split 1.4 1718 1436

  • Fraction of cold leg area (CD for DECLG)

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{

a oc m

o 100 6

a m

a

ss 50 0

O 50 1610 100 200 2!,0 Jus TlWE (seconds)

Figure 1216. Pumped Safety Injection thw for CD=1.0 Split 12 23 i

NSD SAE ESI 97 647 40

-f 35 I

f k

30 -

1 EU E

20 15 10 0

50 100 150 200 250 300 Time (sec)

F Figure 1217. Containment Backpressure for CD=1.0 Split 12 24

=.:-----------

" - - - - -

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