ML19220A509
| ML19220A509 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 04/24/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML19220A498 | List: |
| References | |
| ZAR-780424, NUDOCS 7904230039 | |
| Download: ML19220A509 (15) | |
Text
G 4
ANALY3IS CT SMALL 3REAKS IN THE
?.EACTCR CCCLANT FUMP DISCHARGE PIPING FOR THE 35W LCWERED LCCP 177 FA PLANTS 79042300M APRIL 24, 1973 US 153
g 1.
Introduction On April 14,1978, 3&W reported that previous s=all break analyses had not been based en the worst break locatien.
This report indicated that the worst case break had ccv been deter =ined to be at the reactor coolant pu=p discharge. A spectru= of small braaks has been exs=ined for the B&'J 177-IA lowered loop plants using the s=all break evaluation =cdel described in BAW-10104, Rev 3, "3&W's ECCS Ivaluation Model." These results sh i that it is necessary to use operator actica during the early stages of the pos-tulated accident, to effectively =itigate the accident consequences and
=eet the criteria of 10 CFR 50.46.
Operator actica is used to achieve suf-ficient and balanced flew through all four H?I injection lines.
This re-port shews that operatica up to at least 2368 MUt is possible within the criteria of 10 CFR 50.46 and Appendix K.
2.
Evaluation 2.1.
Method of Analysis The analysis =ethod used for this evaluation is that described in Chapter 5 of BAW-10104, Rev 3, "3&W's ECCS Evaluation Model." Specifically, the model, except for break size, break locatica, and core pouer, is the sane as utilized in Appendix C of BAW-10103A, Rev 3, "ECCS Analysis of B&W's 177-FA Lowered-Loop NSS."
The analysis uses the CRAFT 2 code to develop the history of the reactor coolant systes hydrodyna=ics.
The CRAFT codel uses 19 nodes to si=ulate the reactor coolant syste=, two nodes for the secondary systes, and one node for _the reactor building.
A sche =atic diagrem of the =edel is shown in Figure 1 aleng with the node descriptiens.
Control volumes (noces) in and around the vessel are all conr.ected by a pair of flow paths to permit counter-current fleu. The break is asse=ed to be located at the bottom of the cold leg piping between the reactor coolant pu=p discharge and the reactor vessel. The Wilson, Grenda and Patterson average bubble rise model is used for all nodes. Uithin the core region, however, a cultiplier of 2.33 is applied to the calculated bubble rise velocity Appendix ? of 3A'.i-10104 demonstrates that a cultiplier of 2.33 in CRAFT 2 gives a mixture height within 12% of that predicted by FOAM.
Thus, no F0AM analysis aill be needed if the CRAFT 2 eixture level rc ains above the core by 21 of the active lengen.
The follouing ascurptiens are made for ecuditicas and systec responses during the accident : Ur154
a.
The reactor is operating at 102% of the steady-state power level of 2568 MWt.
For breaks greater than 0.1 f t, the analysis utili:ed a power level of 102% of 2772 MNt.
b.
The leak occurs instantaneously, and a discharge coefficient of 1.0
/
is used for the entire analysis. Bernoulli's equatica was used for the subcooled portion of the transient, while hbody's c.orrelation was used in the two-phase portion.
c.
No offsite power is available.
d.
The reactor trips on icv pressure at 1900 psia.
e.
The safety rods begin entering the core af ter a 0.5 second delay from the ti=e the reactor trip signal is reached.
f.
The RC pu=ps trip and coast down coincident with reactor trip.
g.
One co=plete train of the emergency safeguards system fails to operate, leaving two CFTs and only one HPI and one LPI system available for pumped injection to mitigate the consequences of a cold leg break, h.
The auxiliary feedwater (FW) system is assumed to be available during the transient. Its = sin function is to re=ove heat from the upper half of the steam generator during 1 1:.itial stages of the transient.
When the secondary side of the steas generator beco=es a source of heat to the pri=ary system, the assu=ption of auxiliary IW maxi =1:es the energy that =ust be relieved.
1.
ESFAS signal error band is considered in the analysis to signal the actuation of the EPI system.
- j. The peak linear heat generation rate in the hot pin is the maxi =u=
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 =inutes following the ECCS initiation signal.
This assu=p-tion is explained = ore fully belou and in section 3.
As =ost of the breaks evaluated in this spectrum showed core uncovery, te=perature calculations were necessary.
Once core uncovery occurs a spatial swell distribution aa. lysis is necessary to _ assura that only the core covered by =ia:ure is included in the swell level.
35U user
-"e FO.U!
code.
The code was utilized under the same assu=ptions as dese:
J above with the follcwing additions:
iib' 155 1.
The pruer shape shown in Figure 6 was used but i=ple=ented with a radial peaking factor of 1.C.
This represents the aver:ge channel condition which is appropriate for use in swell level calculations.
2.
Stea= production due to heat fro = the pri=ary metal, core and lever plenu= flashing, was conservatively underpredicted. Although the CRAFT codel accurately predicted these effects, full credit was not included in the FOAM simulation as a conservative ce=putational con-venience. This si=ulation, therefore, underpredicts both the swell level and the stea=ing rate.
Consequently, = ore core uncovery and lower coolant flow are used in the heat-up evaluation.
The heat-up calculation was perfor=ed using the THETA code in the =anner assu=p tions described in section 5 of BAW-10104. The following addi
.a are utilized in the THETA evaluation:
1.
The power shape of Figura 5 was used with a radial power factor of 1.8.
This =ax1=1:es stea= superheating,and sets the peak local power at 10.5 f t at the technical specification LOCA li=it.
2.
Coolant flow and =1xture level were taken direct.'y fro = the F0AM calcu-lations.
3.
End of life pin pressures were used to conservatively predict the inci-dence of fuel pin rupture.
2.2.
High Pressure Injection Syste= Performance The previous arrangement of the HPI syste= allowed for one pt=p to inject into the reactor coolant syste= (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 flow 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 firw would be lost.
Since the flow fro = one H?I pu=p will nou be distributed to four injection poin:a and to assure conservatis= in allouing for injection line loss dif-farences, this analysis assu=es 20% of the HPI is injected into the broken cold leg.
The imple=ented action starts at 5 minutes af ter an ECCS signal
~
and is concluded 15 ninutes af ter the signcl.
The resultant HPI flow can be conservatively represented as a linear ramp froa 5 to 15 minutes.
This ramp 6w 156 was si=ulated in our present CRAFT code as a step function at 650 seconds (600 seconds for action, 30 seconds for ECCS signal).
This is illustrated in Figure 7.
2.3. Break Soectrue and Results All evaluatiens reported in this analysis assu=a the high pressure injection performances as described in section 2.2.
Breaks of 0.3, 0.2, 0.13, 0.1, 2
2 0.07, and 0.04 ft were evaluated. The evaluation of a 0.5 ft break was repcrted in BNJ-10103A, Rev 3, and shows co=plete core covery at all ti=es and thus no temperature 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 flew within 2000 seconds except for the 0.04 ft break.
Figure 3 shows (CRAFT) =ixture height as a function of ti=e for each break of the spectru=. As can be seen, breaks of approximately 0.3 f t and larger than approxi=stely 0.04 ft uncover part of the core.
Various uncovery levels and ti=es are observed but all trends are consistent throughout the spectru=.
The 0.04 ft break achieves a =atch up of effective ECCS (the RPL injected into the intact cold legs) with the core decay heat and the RCS metal heat at 2500 seconds. After 2500 seconds the
~4
"-a level will rise in the core due to excess EPI injection. As the 0.04 ft break has a level of 14 feet at this ci=e the core never uncovers and no.te=perature excursion occurs.
For breaks s= aller than 0.04, the match up will occur at approximately the same ti=a and the core =ixture levels will drop slower; thus, for all smaller breaks the core will re=ain covered.
Figure 4 shows the ti=e duration of uncovery for four core elevations as a function of break size. These results are from CRAFT.
As can be seen, the
=axi=u: degree of uncovery and the =aximu= ti=c of uncovery occur for the 0.15 ft break and is the worst case break.
This break can thus be identi-fled as the worst case. A si=ilar uncovery occurs for the 0.1 and 0.07 ft 9
breaks.
The 0.07, 0.10, and 0.15 f t' breaks.cre analyzed for te=perature recpence.
T'ce 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 spec-tru: are within acceptance criteria.
~
9 The 0.2 ft break was not evaluated for three reascas:
1.
The uncovery ti=e is approximately 2/3 of that for the 0.15 ft break.
2.
The depth of uncovery is only 1/2 of that for the 0.15 f t b reak. The 2
=ini=us core level is only 11 feet as co= pared to 10 feet for the 0.15 ft-break.
3.
The decay heat rate will be approxi=stely the sa=e as for the 0.15 f t b reak.
Thus, the 0.2 ft case is well bounded by the 0.15 f t case.
Local =etal water reactica is shewn below the te=perature cur cas on Figure 5.
The highest value is 2.S~ for the 0.15 ft break.
This value is well below the local oxidation 11=it for the large breaks utilized in BAN-10103 for the whole-core =etal-water reaction calculation. Thus, the whole-core =etal-water reaction results given in section 8 of BAW-10103 is conservative for small breaks. The degree of clad da= age is bounded by the large break results which produce higher clad te=peratures. Thus, all criteria of 10 CFR 50.46 are
=et.
This analysis is conservative for =any reasons as detailed in the write-up and =cets all evaluation criteria. This analysis shows that all 177 lowered loop plants =eet the criteria of 10 CFR 50.46 if operated at or below 2568 MWt power and in conjunction with the specif1ed operator action.
3.
Operator Action The ECCS analysis used as a basis for this report assu=es that the operative HPI train (one train is. lost due to a single active failure) provides e=er-gency core cooling water to the RC loop containing the break.
It is conser-vatively assu=ed that the break is on the lower portion of the reactor cool-ant pump discharge piping resulting in the total loss to the syste= of 50%
of the available HPI flow. Acceptable =itigation of the accident requires
= ore than the 50 % of this flew from one RF1 purp.
If, following the LCCA, it is assu=ed that one train of HPI does not start it is necessary to take operator action to achieve a flow split "herein no more than 30% of the re-
=aining purp's flow goas into the cold leg centaining the break.
The follou-ing is a description of the action required for a typical plant.
~
1.
Upon ESFAS signci check for flow throu;;h both HPI trains.
La 158 2.
If no flow in one train:
-- open pump header cross-connect valves
-- check. EFI valve position and open il closed 3.
Secure flew through nor=al makeup line if flov is indicated 4.
Throttle RPI valves as required to balance flow and =eet run out li=its The above actions initiated at five sinutes and completed within 15 minutes subsequent to the ESTAS actuation ensures adequate HPI ficw 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 time for a 10 =inute 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 si=ulated by a step function increase, 10 minutes after ESTAS actuation.
4.
Evaluation of Other B&W Sunolied Plants a.
Davis-Besse -- The DB-1, 2 and 3 Plants have been analyzed for a spectrum of small breaks at the RCP discharge in accordance with an approved small break evaluation =odel.
This analysis is reported in BAU-10075A, Rev 1, March 1976.
In addition, the Davis-Besse 1, 2, and 3 units have a split high pressure injection and cakeup system design.
The Davis-Besse RPI pu=ps, therefore, have considerably higher capacity at the system pres-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 BAU-10062A, Rev 1, March 1976.
In addition, the 205 and 145 FA HPI sys-tens contain cross connects between the two IIPI trains dormstream of the HPI injection valves.
These cross connects effectively achieve the same flow split as the operator action assured in the current 177 FA lowered loop analysis and the flow split is achieved when the HPI pump is started.
65 139 All BSU supplied plants except the 177 FA lcwered 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 boilof f.
9 ere gb-160
.h Figure
-1.
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FIGURE 5.
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