ML19259B296
| ML19259B296 | |
| Person / Time | |
|---|---|
| Site: | La Crosse File:Dairyland Power Cooperative icon.png |
| Issue date: | 01/11/1979 |
| From: | Ziemann D Office of Nuclear Reactor Regulation |
| To: | Linder F DAIRYLAND POWER COOPERATIVE |
| References | |
| NUDOCS 7901260021 | |
| Download: ML19259B296 (7) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION f~[ h WASHINGTON. D. C. 20555 i':Mk//
January 11, 1979 p
Docket !!o. 50-409 Mr. Frank Linder General Manager Dairyland Power Cooperative 2615 East Avenue, South La Crosse, Wisconsin 54601
Dear Mr. Linder:
Enclosed is a copy of our redraft evaluation of Systematic Evaluation Program topic IV-1 A (Operation with less than all loops in service).
You ',r3 requested to examine the facts upon which the staff has based it, evaluation and respond either by confirming that the facts are e,orrect, or by identifying any error.
If in error, please supply corrected information for the docket. We encourage you to supply for the docket any other material related to this topic that might affect the staff's evaluation.
Please note that this evaluation supersedes the evaluation issued by our letter dated August 17, 1978.
It would be most helpful if your comments were received within 30 days of the date you receive this letter.
Sincerely, 2,414 %
Dennis L. Ziemann hief Operating Reactors Branch #2 Division of Operating Reactors
Enclosures:
Topics IV-1A cc w/ enclosures:
See next page G
79012600 2 (
f
- *. F r 3 r.L '. i nd e r January 11, 1979 CC Fritz Schubart, Esquire Staff Attorney Dairyland Power Cooperative 2615 East Avenue South La Crosse, Wisconsin 54601
- 0. S. Heistand, Jr., Escuire Morgan, Lewis & Bockius 1800 M Street, M. W.
Washington, D. C.
20036 Mr. R. E. Shimshak La Crosse Eoiling Water Reactor Dairyland Powe-Cooperative
?. O. Box 135 Genoa, Uisconsin 54632 Coulee Region Energy Coalition ATT';:
George R.
Nygaard P. O. Eox 1583 La Crosse, Wisconsin 64601 La Crcsse anlic L;brary 800 Main Stree:
La Crosse, Wisconsin 54601 K M C Inc.
ATTN:
Mr. Jack McEwen 1747 Pennsylvania Avenue, N. W.
Suite 1050 Washington, D. C.
20006
SYSTE".ATIC EVALUATIO" PROGRAM Tcpic TV-I-A:
Optra-ion with less than all locps in service.
PLANT: Lacrosse 5:iling Water Reactor (LACSWR)
Discussion The majority of the prisently operating BWR's and PWR's are designid to pernit opere-ion with less than full reactor coolant flow.
If a P'.;R reacter c0:lant puro or a BWR recirculation purp becomes inoperati.e, the fic'.. provided by the remaining loop or loops is sufficient for steacy state o era-ion at some definable power level, usually less "ir
~.11 co.ce.
Plants authorized for lona term ooeration with one reactor coolant :umo or recirculation oumo out'of service have submitted, and the staff.as accroved, the necessary ECCS, steady state, and transient analyses.
The rerainino PWR and BWR licensees have Technical Snecifications which require reactor shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> if one of the operating loons becones inoperable and cannot be returned to operation within the tire period.
LACBWR has n'o restriction on ooerating in the (n-1) 1000 node and our review indicated that analyses to justify ooeration in this mode had not been submitted.
Evaluation Several factors have to be considered wnen evalua*.ing (n-1) loop (1) the impact on normal operation (i.e. are there adec. ate
- eration; thermal margins when one considers the effect of anticipated trans-
- ents),
(2) the potential for a new accident (in this case a coldwater accident caused by the startup of the inactive pump), and (3) the potential effect on accidents which are analyzed (principally the LOCA and locked r: tor accident).
that can affect all three of these considerations is thi effect One factc e LACBWR is a of one loop operation on reactor coolant flow distribution.
The forced-circulation two loop BWR with a thermal rating of 165 megawatts.
The system cools.'the reactor by circulating as much.-as 30,000 gpm of water.
primary water flows upward through the core and then downward through the The water then steam separators to the recirculating water outlet plenum.
flows to the 16 inch forced-circulation pump suction manifold through four 16 inch no:zles and is mixed with reactor feedwater that enters the manifold from four 4 inch connections.
From the manifold, the water flows -hrough 20-inch suction lines to the two 15,000 gpm variable speed forced-:ircu-Hydraulically-operated rotoport valves are at the suction la:icn pumps.
The 20 inch pump discharge line returns' the and discharge of each pump.
From water to the 16 in:n forced circulation pump discharge manifold.
tne manifold, the water flows back to the core through four equally spaced iE irch nozzles in the bottom of the reactor vessel.
?.
The :hysical arrancerant of the forced circulation system as :sscribed above li.e. cc~ :n'c.ischarge header and com cn suction header'i intuitively indicates that flow perturbations will not be in: reduced to the system, that is, at a specified f?ow it is expected that the reactor i
will not oiscern the difference between two put:s or cne pumo c;erating,
, j n a fdi ti c r., a fl:rt stability analysis, performed by Allis-Chalmers,,
(reference 2), simula:ing cne pum: and two pump flow situations at LACE'.G shows that no flow oscillations are expected to occur during one pump cperation.
The staff concluces that due to the system con-figuratien and the Allis-Chalmers analysis uneven or asyre:ric flow tondi:icns will be neg.lible.
'li:n regard to r.ormal operation (with one loop), since we have con-ciuted :Sa: there are minimal effects on flow cistribution, One thermal ar cin pro:ec:ico specified in the Technical Specification fer recucec ficw are adequa:e.
.,i:n e; arc tc :ne pt:ential for a new accicent, the staff considered ne to:ential fcr a cold water injection accident caused by the startup of the inactive 1o00.
The LACB'.2 forced-circulaticn s'ystem 20 inch c;s:harge va've is electrically i :sric9ed t; ci:se v.henever its ass:ciated prp is tri;;e: cr a: zero s;sec and ;he pur: in the other loop is at greater than zero speed.
A second intericck :ontrols the position of a 2 inch bypass valve around the cischarge valve for the rurpose of maintaining a thermal epuilibrium
- i: ser :He :..: 1:::s.
An interlock causes the dischar:e.'alve ::
's ain in the ciesed positien anvtime the temcerature cifferen:ial be: ween One iceps is grea:er than 10 F, 'The bypass valve opens anytime the dis-0 charge valve is out of the fully open position.
Durin0 single loop c:eration the suction valve is locked in the fully open position.
This configuration provides for a backflow through the loop which maintains the temperature differential at a minimum.
We conclude, that since the
- ~ idle icop is maintained in tehrmal equilibrium by interlock with the operating loop, ne significant cold water injection can. occur and therefore cdditional restrictions are not required on reinstating the idle loop to operation.
With regard,to the potential effect on postulated accidents, the staff considers the LOCA and locked rotor to be the most bounding in terms of the effects of single loop operation.
For the LOCA the primary
3 areas f con:ern are (1) the effect of difference in water inve.: cry (in :he vessel and cperating loops), ('2) energy removal fror the core C.:ri".g. blowd:wn due to ;ath of primary water to the break, and
'(3) tre effe:t of reduced power level and stored energy in the fuel.
Re;a-din; the viater inv=ntory, the reactor water level control system a: '.1 :EW: is one of essentially c:nstant mass, that is, water level is ir.:rease: or decreased to account for the steam voids presen; durin; varia:ic.s in power level.
Although the indicated water level at Li EW: is a reasure of the actual inventory in the reactor vessel, sir.ce the lexer nozzle for water level measurement is below the regions ir : 5 :: e : hat cer.:ain significant voids, the level cro;ra~ is t!!s:
cn stia
<: ids present during full power and full flow.
Since :his ::de C
- eration yields essentially the same void fraction a: 103 cercent
'i :.. S r. : 50 ;ercen: ' low, the inventory of water in the vessel does not c h a r. ;e.
The staff, therefore concludes, that it is reasonable to assume t.a: Pere are no significant inventory differences between single and cual ico: operation.
Further assurance of maintaining constan mass is crovided by a Technical Specification (4.2.2.9) limiting operation to 5: per:er; of rated reactor power during single ices
- eration.
Tnis b unes LI:5WR so that control rods cannot be varried to increase reactor
- 'er v."le at maxirr flow on one pump thereby producing a higher reactor a:<;er res iting in a greater void fraction and potential for decreasing the ras s inventory.
Additionally, since the bypass line is oper, in tne i:le 10:p (discussec above), the inventory of water is essentially the same as : a: used in the '_0C A analysis.
The staff therefers concludes, based en
- ne a::ve, that the inver::r availabie curir.g sir;1e '::: ::s a:i:
w:uic r.:: te a;oreciable changed from that assumed available during duai iocp opera:icn at full power and full flow.
Regarding the effect on energy removal due to blowdown flow distribution,
,the location of the limiting LOCA is such that blowdown effluent would be preferentialy routed through the core for that portion of the transient before the 2" bypass line is closed and the 20" re:irc line opened. The change in blowdown flow will be slight but in a direction to enhance heat removal from the core.
The stored energy and decay heat strongly affect peak clad temperatures.
If the power level is effectively halved, the stored energy of the fuel is crocortionally reduced resulting in peak clad temperatures signifi-
- a.f y below those calculated.
Fuel burnup and the clad gap effect the reia-icr. snip between power level and stored er. erg., ir :he fuel bu: are
. seccndary effects so that reducing power level by a factor cf 2 reduces the stored energy by approximately 2 with a corresponding decrease ir peak ciad terpera ure during the LOCA.
The a::ident m:st affe:ted by the single loop operating mode is the ic:ked cump rotor.
A seizure of the rotor in the operating loop pumo causes a complete loss of pum:ing flow without benefit of pung coas -
d : '. n.
- n the even: of disrupted flow at. LACSWR, reactor water level wcuid ce:rease to the reactor scram se point and initiate Ine high pressure em5cgency c re so'ay.
This sare signal would cause the reactor building stea-is:la-ion valve to close.
Closure of the steam isolation valve "i-t a:es s t.;-d... conder.ser coerati:r..
The forced circulatior sys er dis "I';e an: su:-ion valves are interlocked with the shutd:wn con-denser is in service providing a flow path for natural circulation.
' nu:fowr. ccoling by natural circulation through the shutdown conderser
<.:.;1: Or::eed in a normal nanr.er exce:: that flow resistance w:uld De greater in one 100.
We estimate tha-flow resistance (K) to be at:ut
- times higher with a locked rotor than a free wheeling pum? based on LOFT precictions.
Paragraph 14.3.9 of LCBWR Safeguards report states that na ural circulation flow (estimated to be 4C00 gpm) can remove up
- 23.cer:ent of full power heat' generation at 1235 psig without violating therral-hydraulic design criteria or a;proaching burnout heat fiux.
' C F.;
is res ric ed :: O pcv.er by tech spec when in on n-1 opera-ing mcde.
W.en reactor scram occurs, the power generation will im ediafely de:rease tc about 75 of initial power or 3.5!; of full povier.
heat
- eneraticn w
- uld then cecay, folle ir; the A"S curve.
Ta:le 14-3A of the Safeguards report indicates that 2700 gpm na urai circula-ion will remove 65; power.
Our conclusion is that even with the flow resistance increase in one recirculation loop, decay heat could be r.emoved by natural circulation alone.
Additional assurance i.s derived by actuation of the high pressure core' spray and operation of the Shutdown condenser following a locked rotor accident.
Based on our review of the above we conclude that operation with less than all loops in service at LACBWR is acceptable and no facility or Technical Specification changes are required.
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