ML19220A519
| ML19220A519 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 05/01/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19220A498 | List: |
| References | |
| ZAR-780501, NUDOCS 7904230043 | |
| Download: ML19220A519 (16) | |
Text
ANALYSIS OF S193.L BREARS Ili THE REACTOR COOLANT PRIP DISCHARGE PIPING FOR T3E E ' LOWERED LOCJ 177 FA PLMITS
}!ay 1, 1978 7904230o41 W.
GL-1ES
h 1.
Introduction On April 14, 1978, B&W reported that previous s=all break analyses had not been based en the worst break location.
This report indicated that the vorst case break had now been determined to be at the reactor coolant pump discharge. A spectrum of s=all breaks has been examined for the B&W 177-FA lowered loop plants using the small break evaluation model described in BAW-10104, Rev 3, "B&W's ECCS Evaluation Model." These results show that it is necessary to use operator action during the early stages of the pos-tulated accident, to effectively citigate the accident consequences and meet the criteria of 10 CFR 50.46.
Operator action is used to achieve suf-ficient and balanced flow through all four HPI injection lines. This re-port shows that operation up to at least 2568 MWe is possible within the criteria of 10 CFR 50.46 and Appendix K.
2.
Evaluation 2.1.
Method of Analysis The analysis method used for this evaluation is that described in Chapter 5 of BAR-10104, Rev 3, "B&W's ECCS Evaluation Model." Specifically, the model, except for break size, break location, and core power, is the sa=e as utilized in Appendix C of BAW-10103A, Rev 3, "ECCS Analysis of B&W's 177-FA Lowered-Loop USS."
The analysis uses the CRAFT 2 code to develop the history of the reactor coolant system hydrodynamics.
The CRAFT odel uses 19 nodes to si=ulate the reactor coolant system, two nodes for the secondary system, and one code for the reactor building.
A sche =atic diagram of the codel is shown in Figure 1 along with the node descriptions.
Control volu=es (nodes) in and around the vessel are all connected by a pair of flow psths to permit counter-current flow. The break is~ assumed to be located at the bottom of the cold leg piping between the reactor coolant pu=p discharge and the reactor vessel.
The Uilson, Grenda and Pattersen average bubble rise codel is used for ali nodes. Within the core regica, however, a cultiplier of 2.38 is applied to the calculated _
bubble rise velocity.
Appendix F of BAW-10104 demonstrates that a v a.t>
! 2.?R in CRAFT 2 gives a mixture height within +2: of that c FOAM.
Thus, no FOAM analysis will be needed if the CRAFT 2 m.
-' lavel remains above the core by 2 of the active length.
The following assumptions are made for conditions and system responsef3,'f'3Jbi) durine the accident:
The reactor is ope.ating at 1021 of the steady-state power level of a.
2568 MWt.
b.
Ihe leak occurs instantaneously, and a discharge coefficient of 1.0 is used for the entire analysis. Bernoulli's equatien was used for the subcooled portion of thu transient, while Moody's correlation was used in the two-phase portion.
c.
No offsite power is available.
d.
The reactor trips on low pressure at 1900 psia.
e.
The safety rods begin entering the core after a 0.5 second delay from the ti=e the reactor trip signal is reached.
f.
Tne RC pu=ps trip and coast down coincident with reactor trip.
g.
One co=plete train of the e:ergency safeguards system fails to operate, leaving two CFIs and only one HP1 and one LPI system available for pu= ped injection to mitigate the consequences of a cold leg break.
b.
The auxiliary feedwater (W) system is assu=ed to be available during the transient.
Its main function is to remove heat from the upper half of the steam generator during the initial stages of the transient.
When the secondary side of the steam generator becomes a source of heat to the primary system, the assu=ption of auxiliary W =axi=1:es the energy that must be relieved.
1.
ESFAS signal error band is considered in the analysis to signal the actuation of the HPI system.
- j. The peak linear heat generation rate in the hot pin is the maximum allowed by the Technical Specifications at the 10.5 ft level.
k.
Operator action is taken to increase the HPI flows to the intact cold legs at 10 minutes following the ECCS initiation signal. This assu. p-tion is explaiced more fully below and in section 3.
As most of the breaks evaluated in this spectrum showed core uncoverf, temperature calculations were necessary. Once core uncovery occurs a spatial swell distribution analysis is necessary to assure that only the core covered by mixture is included in the swell level. B&W uses the F0/JI code. The code was utilized under the same assumptions as described above with the following ' additions: US 170
1.
The power shape shown in Figure 6 was used but i=pleranted with a radial peaking factor of 1.0.
This represents the average channel condition which is appropriate for use in swell level calculations.
2.
Steam production due to heat from the pri=ary uetal, core and lower P enum flashing, was conservatively underpredicted. Although the l
CRAFT model accurately predicted these effr. cts, full credit was not included in the FOAM si=ulation as a conservative computational con-venience. This simulation, therefore, underpredicts both the swell level and the steaming race.
Consequently, = ore core uncovery and lower coolant flow are used in the heat-up evaluation.
The heat-up calculation was performed using the THETA code in the manner described in section 5 of 3rn'-10104.
The following additional assu=ptions are utili=ed la the THETA evaluation:
1.
The power shape of Figure 5 was used with a radial power factor of 1.8.
This caximizes steam superheating and sets the peak local power at 10.5 ft at the technical specification LOCA limit.
2.
Coolant flow and mixture level were taken directly from the FOAM calcu-lations.
3.
End of life pin pressures were used to conservatively predict the inci-dence of fuel pin rupture.
2.2.
H12h Pressure Injection System Perfor=ance The previous arrange =ent of the HPI system allowed for one pu=p to inject into the reactor coolant system (RCS) at two locations. As one injection Point could be in the region of the break, 50% of the one HPI flow could fail to penetrate the reactor vessel. This ficw would, therefore, not be avail-able to provide core cooling. The proposed operator action, section 3, will provide four points of penetration of the RCS. Therefore, only 25% of the HPI flow would be loct.
Since the flow from one HPI pu=p will now be distributed to four injection points and to assure conservatism in allowing for injection line loss dif-ferences, this analysis assu=es 30% of the HPI is injected into the broken cold leg. The implemented action starts at 5 minutes after an ECCS signal and is concluded 15 minutes af ter the signal.
The resultant HPI flow can be conservatively represented as a linear ramp from 5 to 15 minutes.
This ramp ou 171
vas si=ulated in our present CRAFT code as a step function at 650 seconds (600 seconds for action, 50 seconds for ECCS signal). This is illustrated in Figure 7.
2.3.
Break Soectrum and Results, All evaluations reported in this analysis assu=e the high pressure injection perfor=ance as described in section 2.2.
Breaks of 0.04, 0.07, 0.1, 0.13, 0.15, and 0.17 ft were evaluated. The evaluation of a 0.5 ft break was reported in 3AW-10102A, Rev 3, and shows co=plete core covery at all times and thus no te=perature excursion. The 0.5 ft break results are independent of HPI flow and re=ain valid.
Figure 2 shows the RCS pressure transient for each break. As shown, each ac-cident initiates CFT flow within 2000 seconds except for the 0.04 ft break.
Figure 3 shows (CRAFT) =ixture height as a function of time for each break of the spectru=. Various uncovery levels and times are observed but all trends are consistent throughout the spectru=.
The 0.04 ft break achieves a =atch up of effective ECCS (the HPI injected into the intact cold legs) with the core decay heat and the RCS =etal heat at 2500 seconds. After 2500 seconds the =ixture level vill rise in the core due to excess HFI injection. As the 0.04 ft break has a level of 14 feet at this ti=e the core never uncovers and no te=perature excursion occurs.
For breaks s= aller than 0.04, the =atch up will occur at approxi=ately the sa=e time and the core =ixture levels will drop slower; thus, for all s= aller breaks the core will re=ain covered.
Figure 4 shows the time duration of uncovery for three core elevations as a function of break si:c. These results are fro: C2 AFT. As can be seen, the
=axi=u degree of uncovery and the =axi=uu time of uncovery occur for the 0.13 ft break and is the worst case break in this analysis. This break can thus be identified as the worst case for operation up to 2568 FMt.
The 0.07, 0.10, 0.13, 0.15, and 0.17 ft breaks were analyzed for te=perature response.
The results are shown in Figure 5 and are well within the criteria of 10 CFR 50.46. They provide positive assurance that all breaks of the spectru are within acceptance criteria.
The evaluation of break sizes of 0.2 and 0.3 ft was reported in the report of April 25,1978 (J.H. Taylor to R.L. Baer) for a power level of 2772 FNt and is thus conservative for the analysis herein.
bTM Local ectal-water reaction is shown in Table 1.
The highest value is 1.72%
for the 0.13 ft break. This value is well below the local oxidation limit for the large breaks utilized in BAW-10103 for the whole-core metal-water re-cetion calculation. Thus, the whole-core metal-water reaction results given in section 8 of BAW-10103 is conservative for s=all breaks. The degree of clad da= age is bounded by the large break results which produce higher clad temperatures. Thus, all criteria of 10 CFR 50.46 are met. This analysis is conservative for many reasons as detailed in the writeup and =eets all evalu-ation criteria. This analysis shows that all 177 lowered loop plants meet the criteria of 10 CFR 50.46 if operated at or below 2568 MWt power and in conjunction with the specified operator action.
3.
Operator Action The ECCS analysis used as a basis for this report assu=es that the operative EPI train (one train is lost due to a single active failure) provides ecer-gency core cooling water to the RC loop containing the break.
It is conser-vatively assumed that the break is on the lower portion of the reactor cool-ant pu=p discharge piping resulting in the total loss to the syste= of 50%
of the available HPI flow. Acceptable citigation of the accident requires more than the 50 % of this flow from one HPI pu=p.
If, following the LOCA, it is assumed that one train of HPI does not start it is necessary to take operator action to achieve a flow split wherein no core than 30% of the re-maining pu=p's flow goes into the cold leg containing the break.
The follow-ing is a description of the action required for a typical plant.
1.
Upon ESEAS signal check for flow through both HPI trains.
2.
If no flow in one train:
-- open pu=p header cross-connect valves
-- check HPI valve position and open if closed 3.
Secure flow through normal makeup line if flow is indicated 4.
Throttle HPI valves as required to balance flev and meet run out limits e
f173
_5-
The above actions initiated at five =inutes and completed within 15 ninutes subsequent to the ESFAS actuation ensures adequate HPI flow for accident =iti-gation.
In the analysis, credit is taken for the HPI flow as the HPI injec-tion valves are opened. Figure 7 shows the calculated HPI flow for a typical plant as a function of ti=e for a 10 minute valve opening. As shown in Figure 7, the =ajority of the HPI pu=p capacity would be delivered with a partial valve opening. For the small break analysis, a linear flow versus valve posi-tion response was simulated by a step function increase,10 minutes after ESFAS actuation.
4.
Evaluation of Other B&W Sucolied Plants Davis-Besse -- The D3-1, 2 and 3 Plants have been analyzed for a spectrum a.
of s=all breaks at the RCP discharge in accordance with an approved small break evalu tion =odel.
This analysis is reported in BAW-10075A, Rev 1, March 1976.
In addition, the Davis-3 esse 1, 2, and 3 units have a split high pressure injection.nd =akeup syste= design. The Davis-Besse HPI therefore, have considerably higher capacity at the system pres-pu=ps, sures experienced.
b.
205 and 145 FA -- These plants have been analyzed for a spectrum of small breaks at the RCP discharge in accordance with an approved small break evaluation model.
These analyses are reported in BAW-10074A, Rev 1, and
'W-10062A, Rev 1, March 1976.
In addition, the 205 and 145 FA HPI sys-tens contain cross connects between the two HPI trains downstream of the HPI injection valves.
These cross connects effectively achieve the sa=e flow split as the operator action assuned in the current 177 FA lowered loop analysis and the flow split is achieved when the HPI pu=p is started.
All 35W supplied plants except the 177 FA lowered loop plants have raised c.
loops and thus do not trap a large volu=e of coolant in the cold leg.
The raised loop design allows this coolant to drain into the core for core covering and cooling by boiloff.
-'~
Gb 174
TABLE 1 PEAK CLADDING TEMPERATURE VERSUS BREAK SIZE (All at 2568 MNt)
Peak Maximum Time of Break Cladding Local MW Peak Size Temperature Reaction Temperature
[Fta)
(oF)
(%)
(sec.)
0.04 Core Stays Covered With No Temperature Excursion 0.07 1320
.73%
1600 0.1 1440 1.68%
1080 0.13 1551 1.72%
82C s.15 1455 1.67%
740 0.17 1248 0.72%
650 9
~
gry173
FigIl
-1.
CRAIT2 Noding Diagra:n je Sc:all Break Agl 1@l 3
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}1 ode No.
Tdenefficatfed Path No.
_Identificarien 1
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Lover Plenu:2 3,4,18,19 Eat Les Piping 3
Core, Core 3ypass, Upper
$,20 Ect Leg. Upper Plenus, Upper Head 6.21 SC Tubes 4,14 Ect Leg Piping 7.22 SG Lover Eezd 3,15 -
steam Generator Upper 8
Core 3ypass Head, SG Tubes (Upper Ealf) 9,13,24 Cold Leg Piping 6,16 SG Tubes (Lower Ealf) 10,14,25 Pumps 8,18 SG Lover Head 11.12,15,16,25.27 Cold Leg Piping 9.11,29 Cold Leg Piping (Pump Sue:1on) 17,31 Downec=er 10,12.20 Cold Leg Piping (Fucy Discharge) 23 LP1 13 Upper Downce=cr 28,29 Upper Dovnconer
- (Above the ( of Nozzle 3elt) 30 Pressuriser 21 Pressurizer 32 22 Contator.ent Vent Valve 33,34 Lesk & Return Path 33,36 EPI 37 Cootsinment Sprays
~GS 176
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= 3 a. 5 g " c.
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Figure 3 CORE WlXTURE liEICilT (CRAFT)
}
ALL AT 2568 MW 40
- 0.07 085217Y
- 0.04 00523DA f
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10 5
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200 400 600 800 1000 1200 1400 1600 1800 2000 Time, s
Figure 4 DURATION OF UNC0VERY FOR TilREE CORE LEVELS (11.5 FT, 11.0 FT, 10.5 FT)
(NOTE 10.0 FT DOES NOT UNCOVER) 600 ALL AT 2568 MWt h
m a
"a 2;
400
~o 3
11.5 FT s
=
j 200 11.0 F1 11 10.5 FT g
C,,
0 I
k n
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0.05 0.1 0.15 0.2 N
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Break Size, f1 0
0 J
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Figure 5 PEAK CLADDING TEMPERATURE PUMP DISCllARGE SMALL BREAKS
^
Ali. AT 2568 NWt 2000 O
C19 w
f 1500 o
2 0.15FT 2
0.07FT 2
5 1000
[
0.1 FTN g
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,W j
500 W
0 200 400 GOD 800 1000 1200 1400 1000 18CO Time, s J
.i
O FIGURE 6.
AXIAL POWER SHAPE 2.0
~
=
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LA.
2 7
w=
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a
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cz w
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8 10 12 ELEVATIDit til CORE O
g
FIGURE 7.
FLOW VS VALVE POS~~^N (Based on Actual L
. Valve y
Po si tion for Oconee liPI Valves) 100 l
90 CALCULATED l
g CURVE
=
g
{,'j 80 l
70 FLOW RESPONSE N
FOR SMALL OREAK E
60 ANALYSIS d
l l
d 50 E
I 40 l
30 I
20 l
1 10 l
1 0
1 1
I I
I I
I I
I 5
6 7
8 9
10 11 12 13 14 15 TIME, MIN
.)
t
Figure 8 PEAK CLA00 LNG TEfAPERATURE VS DREAK SIZE' 10 ALL AT 2500 IAWt p 16 mj 14
=
! 12 a
?
10 c,
C/!
E h
b 8
cc Q
.5 g
E 4
x E
2 0
0 0.05 0.1 0.15 0.2 2
Break Size, FT G
J
.)
.