ML19268C204
| ML19268C204 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/05/1979 |
| From: | Vollmer R NRC - NRC THREE MILE ISLAND TASK FORCE |
| To: | Scott S NRC OFFICE OF ADMINISTRATION (ADM) |
| References | |
| NUDOCS 8001040717 | |
| Download: ML19268C204 (14) | |
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UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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'DEC 51979 4
Docket No. 50-289 MEMORANDUM FOR: Steve Scott, Acting Chief Distribution Services Branch, ADM FROM:
Richard H. Vollmer Director Three Mile Island Support
SUBJECT:
THREE MILE ISLAND UNIT l' BOARD NOTIFICATION 10 CFR 50.54 REQUEST REGARDING DESIGN ADEQUACY OF BABC0CK AND WILC0X NSSS Please forward the enclosed material to the Three Mile Island Unit 1 Board Notification Service Listings.
The material is an October 25, 1979, letter signed by H. R. Denton and an evaluation of the letter.
Yb1" Richard H. Vollmer, Director Three Mile Island Support
Enclosure:
As stated cc w/ enclosure:
R. Vollmer Document Management Branch o
lotte M. Mulkey G. Mazetis 016-Phillips D. Ross H. Silver 80 01040 ?/7
DISTRIBUTION OF BOARD NOTIFICATION Three Mile Island-1 (Docket No. 50-289)
Dr. J. Carson Mark Mr. William M. Mathis Dr. Dade W. Moeller Dr. David Ukrent" Dr. Milton S. Plesset Mr. Jeremiah J. Ray Dr. Paul G. Shewmon Dr. Chester P. Siess Mr. Myer Bender Dr. Max W. Carhon Mr. Jesse C. Ebersole Mr. Harold Etherington Dr. William Kerr Dr. Stephen Lawroski Dr. Harold 'W. Lewis Theodore A. Adler, Esq.
Ms. Karen Sheldon Ms. Marjorie M. Aamodt Mr. Roger Smith John A. Levin, Esq.
Robert Q. Pollard Chauncey Kepford Jordan D. Cunningham, Esq.
Mr. Steven C. Sholly Ms. Fieda Berryhill~
Ms. Holly S. Keck John E. Minnich Robert L. Knupp, Esq.
Walter W. Cohen Ms. Jane Lee Metropolitan Edison Company Mr. Marvin I. Lewis Mr. Thomas Gerusky Ellyn Weiss, Esq.
Honorable Mark Cohen Karin W. Carter, Esq.
George F.
Trowbridge, Esq.
Dr. Linda W. Little Dr.. Walter H. Jordan Ivan W. Smith, Esq.
EVALUATION OF THE ENCLOSEDPATEP.! At It is the staff's judgment that the enclosed materiai is relevant to the l
scope of the proceejings before the ASLB.in the matter of the restart of Three Mile Island Nuclear Station, Unit 1.
The enclosed material (October 25, 1979 letter, from H. R. Denton to the Tennessee Valley Authority) requests information concerning the sensitivity to feedwater transients of the Bellefonte Nuclear Plar.t now under construction which has a Babcock and Wilcox nuclear steam supf.y system.
Similar letters have been sent to all utilities holding construction permits The for plants with a Babcock and Wilcox nuclear steam supply system.
request for information is directly reh Y to the lov water inventory a,nd the presence of a liquid-vapor interface in the B&W once-through steam generator which closely couples the primary system t; the steam generator conditions with a consequently high sensitivity to,feedwater flow rate perturbations.
The additional information being requested from holders of construction permits will allow a determination by the staff whether it is necessary to halt all or portions of the construction of the affected plants pending the results of an ongoing staff study.
The staff has ai study now underway, using Crystal River, Unit 3 an operating facility with a B&W nuclear steam su: ply system, as a reference facility.
The objective of this study is to ite.tify a::ident sequences 05e p.ea: tor leading to core damage having high frequency c: :are: ::
Sa fety Study (WA5H-1400).
The study uses eve ; :res a : faul tree e :hr.i que s.
2 This assessment to be completed in about 6 months will f::Us or. the risk implications of the sensitivity of the B&W design ano 'n the potential s
interactions arising from the integrated control syster..
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. Primary System Per.turbations incu:eo cy un5:
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I.
Introduction' plants emp1'oy a once through st..m generator (OTSG) design, rather B &'a' than U tude st eam generators whi:h are used in other pressuri:ed water tech steam generator nas approximately 15,000 vertical straight reactors.
' tubes, with the primary coolant entering the top at 603 60S'T and eriting the bottom at about 555'F.
Primary coolant flows down inside the steam generator tubes, while the secondary coolant flows up from the b0ttom on the shel'. side of the OTSG, The secondary :oolant turns to. steam about
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half way ve, with tne remeinin; lengin of :ne stear generat:r being used to l
supernet*. tr.e stear.
spa:e of the.0TSG,
' ne secon ary-side heat transfer coefficient, in the stes:
i is me:n less th:n tr.at in the botto: liqui: se:ti:n.
Tr.is rest.ts in a heat transf er rate.from the primary syste whi:h is quite sensitive to tne liquid If a f ee:.ater in:rease Ever.: c::urs, the ievel in the sten: generators.
This liquid-vapor interface rises, increasing the overall heat transfer.
0 initiates an over:coling decreases tne outlet temperature below 555 F an By contrast, event, which can leac to primary syster :eoressurization.
if a f eed. water decrease event o::ers, the.overall heat transfer ce:reases, the outlet primary temperature increases, and a pressuri:ation transient ensues.
in eitner of tnese :ases, the respor.se of ne primary syste: pressure and,
cressuri:er level to a chan;e in mair, fee:-ater fio rate (:r tat;erature) is :omoara-ively rapi:.
Tnese a:i: primary syster
-ess;*e : nan;es due to
- r.an;es in fee:-ate- :enditi:ns 's snow-e ti. as sys er " sensitivity" and is e
1
design.
.' unique to the B&W OTSG t-Fo110 wing' the incident at Three Mile Island, various a::iens were taken to increase the reliability of the auxild.iry feedwater syster.s and improve plant System modifi:stions to in:rease the reliability Of :ne ATW transient response.
However, use of AFV results may have resuhed in more frequent AFW initiation.
In introduction of cold (100 F vs. 400 F) feedwater into the me 0
This may act to enhan:e system sensitivity.
upper section of the steam generators.
Further' system modifications provide control-grade rea: tor tri;s based on While se:endary system malfun:: ions, such as turdine er feecaater pump trip.
these reactor trips de serve to recu:e. undercoeling feedwater transients by amplify subsequent
, reducing rea::er power promptly following LOMTW, they may
,over:ccl i ng.
~
A reexe.mination was made of small break and 1:ss of fee:wa This resulted in a modification of operator procedures for dealing plants.
with a small break, whi:n include promp RCP trip and raising the water le B th these in the steam generators to (95%) to prom:te naturai circulation.
actions are takt:n when a prescribed low pressure se point is reachec in t rea:ter coolant system and for anti:ipated transients su:h as loss of feedsa these actions may amplify undesirable primary system responses.
In addition to the p:st-TMI changes dis:ussed above, a:tions were also t redu:e the :ha11enges to tne power coerated relief valve (PORV) by_
t:
Knile these p int and lowerin; the hi;r. pressure ree.: tor trip.
tne ?0EV se:
a:*i r.s nave Deen su::essful in reducing the fre:Venty Of ?OEV s
' f have resulted in an increased number of reactor trips. This occurs oecause
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the reactor will' n'ow trip for transients it previously would have ridden through by ICS and' PORY operation.
While The staff is :en:trned by the inherent responsiveness :f S&W OT5G design.
some specific instances are presented in the next se: tion of this paper, the s It is felt that good design practice concerns, are also of a general nature.
and maintenance of the defense-in-depth concept, requires a stable well-be'hav system.
To a large part meticulous operator attenti:n and prompt c,anual action is used' en these plants to compensate for the system sensitivity, rather than any inherent :esign features.
l systa=3
. The staff believes that the general stability f the S&W piant contro OTSG feedwater perturbations be should be improved, and that plant response t:
dampened.
II.-
Re:ent Feedwater Transients the staff met with the S&W licen' sees to discuss recent On August 23, 1979 One aspe:t which is of interest is the relationship of fee: water transients.
In at least one the operator.to the functioning of the main f eedwater system.
instance an operator manually opened.a blo:k vaive in series with a control valv This resulted in an overfeed condition.
(partlyopenbutthoughtto'beclosed).
d to the point where the an several re:ent events the feed flow was reduce Subsecuent over#eed reduced pressure to below rea:t:r tripped on high pressure.
-16;; psi, vnerr HPI was ini-iated, rea:ter :::eant p's.:s
- ripped, and auxiliary tne ::p of
..e stear generators, whi:h in:reased feedwater flow intro:u:ed int:
- ne severi:y of tne :::1down transient.
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n,
'l It ' appear's that in many cases the main feedwater control sys requirements, quickly enough or is not sufficiently stable to meet feedwater ii Rather, the ' system will of ten osci11 ate fro: uncerfeed to overfee One causing a reactor trip and scmetimes a high pressure injectien i
und'esirable element of this lack of stabili:y is that over:coling tra
~
in on the primary side proceed very much like a small break LO;A (
i the operators pressurider level and pressure)1 Thus, for a certain period of t me Th. same may not know whether they are having a LOCA or an over:oolin; l' system. This type of behavior can be initiated by the normal res tor contro failure of a centrol.
was demonstrated by a December.1975 event at 0:: nee, where EST actua: ten.
re: order led to reactor trip, a feedwater transient, and grade T,yg in the A partial list of recent B&W transients and their eff ec'.s is co Appendix to nis report.
Role of the Pressuri:er Level indi:stor III.
A major area of concern arising from the S&W OT5G sensitivi Several B&W feedwater transients have led cf pressurizer level indication.
Most nota:ble was a November 1977 incide loss of pressuri:er level indication.
The arrival at Davis Besse where level indication was los for several minutes.
but rate for this' event appears to be on the order of.1.2 per reactor y d feed-could be on the increase due to the potential for more reactor This is of water transients resulting from post-TMI-2 system modifications.
h b
ating con:ern because an over:coling event could empty the pressu leg wnich may interrupt
- ne p:tential for forming a steam tubble in the h:
The staff f eels :nat natu-ai cir:viation, following RCP tri; :n low :ressure.
- .ase natural circulation are somewha:
tne un:ertainties asso:iated w :n w:
nign f:r an event wi:n a re:prrence it.:erval Of a few years.
1
l Additionally, the' staff believes that good design practice and adhere a.
the defensein-depth concept,'would require that plant operators be aware of A icw-level off-scale reading the reactor's status during expe:ted transients.
on pressuri:er Sevel makes it impossible for the operators to assess system ihventory and more difficult to differentiate between an a :ident and an The staff feels that the frequency with which this excessive cooldown transient.
1 situation oc:urs is undesirable.
~
Some concerns also exist with regard to the operation of the'pressuri:ef hea when less cf level takes place.
Nonsafety grade contr:1 circuitry trips the If these nonsafety grade cutoffs heaters off when pressuri:er level is low.
This situation has should fail, the heaters would be kept on while uncovered.
the potential of overheating the pressurher to the f ailure point, as happened
'.with a test reactor at Idaho Falls.
IV.
Role of 105-MFW The 105 appears to paly a significant role in the p,lant's feedwater response.
However, review of The staff is currently reviewing an FMIA study on the ICS, operatin.; Nperience suggests that the ICS of ten is a contributor to feedwa In some cases the ICS appeared inadequate to provide sufficient transient:.
Some of the utility des:riptions of feedwater plant control and stability.
23,1979) transients (as suranari:ed in the minutes of a meeting on August
%' syste:. The following
'empnasi:ed the role of the operator in operating the.
se:ven:e illustrates the. type of event and system response which the staff
- vid p::entially c::ur.
1.
F.es:::t a '100% power.
F.en:::r trip, from arbitrary :ause (does n:: matter).
2.
Plan; stabili:es in het snut own, f:r a few minutes, nea: rejection 2.
valves).
- ndenser (and/or se::ncary dum:
4.
uvertees Sr
...... y.
pressurizer level shrinks, pressure reaches,160 psi, P.S actuates; RCP tripped; AN on.
(Possible RCP seal failure).
0;erator r.anually controls AN (possibly RFW instene or in addition, if 5.
MN n' t isolated su:h that OTSG 1evel comes up to 95; cf operating range.
This massive addition of :old water may lead to emptying of pressuri:er
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and interruption of natural circulation (or, the het leg may flash due to depressurization and, interrupt natural cir:ulation even if pressurizer does not empty).
HPI delivers cold water,' no heat transfer in OT5G; vaper'from : ore 6.
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leads to system repressuri:ation; steam may cendense or PORY may lih.
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No pump restart :riteria availaole, cir:ulation may not be reestablished..f 7.
It_ appears that an upgraded safety evality ICS, wnich is designed to balan power to OT5G ievel in a better f ashion, could redu:e :ne sensitivity, illustrated in the ab:ve sequence.
Y, Role of ECCS and Auxiliary Feedwa ter I
{t is known that some feedwater transients result 'in over:coling to the extent Traditionally, the operator isolates.
that the HP1 a:tuation setpoint is rea:hed.
4 letdown and turns on an extra makeup pump f:11owin; trip so as to avert thi If this manual a: tion is not performed cuickly enough, or if the a:tuation.
cooldown transient is too ' severe, the HPI set point will be reached and Following pro:edures, the operat:r would then trip all ma-automatically started.
based on the plant sy=pt:rs., If
- 001 ant pumps and utilize recovery pr::edures the in:ident was a:tually a feedwater even; and n:t a small 1.00A, he would When pressure has to tne loss of for:ed :it:uitti:n ;re:etures.
- resumaoiy ;
0 re::vered su:n :na: :ne coc1:n syster nas be::me 5: 7 subcooled, the Operat:r One peccler is the :ifficulty in :iff erantiatin; between a s=a*,1
- an se:ure @!.
. e eum e i
The ::erator v:uld be forced -
'. break LOCA and an 4xcessive feedwater transient.
He ever, felie<ing the sr.all to asse.e a sma11 't.00A until proven otherwise.
break pro:edures n'nd inti:duiir.; : eld auxiliary fee: water, :ey increase the
- nitiation of ATL' at: delivery to the CSTG, severity of an overcooling event.
espe:ially if accompanied by filling to the high levei retvited by new pro-Thus, the AFW cedures (95:) will continue the cocidown and depressuri:ation.
sytte acts to increase the responsiveness of the reactor to feedwater transients where ex:essive cooldown is o::vering.
VI, ten:ivsions The staff believes tnat the :brrer.: B&W plants are Overly res::nsive to feecwater transients be:ause.of the OTSG :esign, ;ress;ri:tr si:in; and PDF3 and high pressure trip set point.
5:=e of tr.e ser.sitivity aise arises from per ur: :i:r.s.
.inar.!;va:ies in the ICS to deal with expe:ted piar.:
Regardiess of the reasons, 5&W plants are currently experie-:ing a number
-The staff of fetawater transients which the staff feels are undesirable.
reduce tne piar.: sensitivity believes that modifications shculd be considered t:
tb taece events and thereby improve the defense -n-deptn whi:h will enhan:e tne safety of the plant.
APPillulX.
((EIMAl[It IRATI5i[flT StNRWlY
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DESCRIPil0il 11tAllSIL;4T DAIE Reactor Trip on 181 1: Pressure - 4 to 3 RCP, A-5/G underfed'-7 i At:ll.I11f 9
8/16/79 (0259 Cit -3 Pressure - 3 RCP - A-5/G underfed - 45% Pws 0/16/19 (1125)
Itcactor Irlp on 188 1:
9 lleactor Trip on liigh Pressure
.3 P.CP - A-S/G undirfed - 481Pwr 8/17/79 (0706) ficactor Trip on 1119: Pressure - 3 RCP - A 5/G underfed - 26%
1 8/17/19 (1825)
Itcactor Irlp on low-Low Level in tiati S/G - 101 Pwr.
8/02/19 (0202)
Turbine Trip - Antic..irlp did not, work - Rx Trip on ill Press -
8/13/79 (1749)
Atto-1 itcactor Irlp on Antl. Irlp (LOIW) - 991 Pwr.
6/11/79 (0333)
Reactor flanually Irlpped when FWPT "lB" Tripp'ed Oconce-1 6/11/19 (0152) fleactor 1 rip on liigli Pressure - feedwater escl11ations - 181 P 5/01/19 (0316)
Rea.tur Trip ois Illgh Pressure - feedwater osclliations - 30%
Oconce-2 6/03/19 (2046) r fleact..r Trip on Antic. Irlp (LOTW) - 1001 Pwr.
7/12/19 (1714)
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1 IREP - INITIhl PLANT STUDY _
We have attemoted to develop a general framework for the conduct o A
risk assessment,of a B&W reactor timed at identifyit.;An abs: lute cetermination of sequen:es relative to the Reactor Safety Study.We have selected Crystal for analysis.
Tr.e at:hitect-engineer risk is not intended.
It began cocnereja) operated by F1:r'i.da Power Corporationfor this Babce:k'and Wilcox operation in March 1977.
The project, as presented in figure i. will require the ')11owing tasks A survey of the LER fileslas now established in 0Rhl, and A0 reports, identify well as the Sandia and Fluor Zion systems interutiens studies to 1.
interactions.and co=non mode frilures which have o::urred in sicilse This survey 'should parallel constrection of sys cc iogi: models an trees sin:e it will ensure that actual experien:e is in:orporated into th assessmen;s performed.
Spe:1fic Event trees for loss.cf-cociant 4::idents and transient c LO As and these will in:1ude a feed-2.
attention will be given to more frequen:
E&W plants and will
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. water transient tree which in:orporates exoerien:e atDr.chasis will be given tow exolore the post-TM] modifications. standing the numan coup sequen:e level.
They will Fault trees for the key systems identified in the event trees. level an 3.
be constru::ed to the componen:
Htcan errors wiii be included as well and electri: power :ensiderations.
Our as the aoility of the coerator to c ;e in tne time sean available.
preliminary opinion is that simplified fault trees will,e required auxiliary feedwa:er and se:endary steam relief, nign pressure emergency core cooling in :ne injection and reci ne following systems:
l i
modes low pressure emergency core :coling in d:tn injection and retir modes, containment spray and containment heat rem val system s:voy of loss of AC power, consioering the 430 and 4150 busse emergency diesel generators, witn lidted analysis of high voltag Separate fault trees ull orobably he re;uired for ECCS and AFW5 initiation logic and the system trees mus in:iuce the contrib yard faults.
from auxiliary systems such as instrument air, ventilatten, compon cooling, etc., and control-incuced f ailures.
This basis will will be permitted provided a written basis is pre is expected from further development of the tree.
An inves-igation of the adequacy of high Oressure-iow pressu 4
wi. :o'-inant se:venees to Analysis Of the chysical :nenomena ass::ia e:0::ain es
' eleases #-:r..ne coa::-::-iate ' elease ca e ntainment. This 5.
in categ:ri:ing releases ir.::
will ai:
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To conduct a progrtm of this magnitude in a short time ;eriod, deleys asse:
isted with acquiring and transferring infernation zust de minimi:ed. Optically, the tYtr.t tret and fault tree analysts sh0ald share a :37.03 lo atien during As the faul: trees ;r:;ress cel>< ine the initial portion of the proje:t.
t:0 logic, however, the analysts should be lo:ated at Or near the site with imme:iate ac:ess te as-built : ravings and ro:ecures as well as a re;reser.tative This will termi: verif t:ati:n Of entineering cf the plant operatiens staff.
and prece: ural details and will minimi:e inferr.atien titns'ar and print re-A::ess shocid aise be arranged between the f 3.J1L tree ar.alysts prede:tien.
4t.the site, the remaining team in Bethesca, the archite:t Engineer, and the vender.
In addition to resi: plant data, deterministic. cal vlatiens may be retuired to This way understand the behavice of the plant ander off-nor:a1 ::ncitiens.
also involve real-tine simuintion at an a;;repriate simulat:r to the extent The arrangements with the vend:r should ever this pessibility and p ssible.
it ray be desirable te have confirmatory :alculati:ns made by one of the NRC Centra: ers en a sele:ted basis-e e
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ENOLOSURE 3 PRELIMINARY IDENTIFICATION OF SYSTEMS AND cogp;NENTS TxAT PAY BE IMPACTE:
EY DESIGN CFdNGES HPI System EFW System DMR System CTi System ROS Pressure Control System Makeup / Letdown System SG Pressure Control System Steam Generator Pressurizer Quench Tank Control Room Layout RCS Piping
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UNITED STATES NUCLEAR REGULATORY CO'.',Mt$$1CN T "I
- T n As motow. o c. aosts i., 4 f f s
oc eber 25, 19-~
% w..'..f Docket he.: 50-438 and 50-439 Mr. H. G. P a rri s
.tnager of power Tennessee Valley Authority.
500 Chestnut Street, iower !!
Chattanooga, Tennessee 27401
Dear Mr. Parris:
SUBJECT:
10 CFR 50.54 REOUEST REGARDING THE DESIGN ADE00ACY OF BABCOCX
& WILC0x NUCLEAR STEAM SUPPLY SYSTEMS UT:LI2:N3 CN"E THROUGH STEAM GENERATORS (BELLEFONTE NUCLEAR PLANT)
Several har: ware and pro:ecural changes have been cace :: c:erating B&W plants
- recu:e the likelihood of recurrence of a TMI-ty:e a::icent. These changes have been in the area of auxiliary fee: water syste s, v.tr; rated control system, reactor protection system, small-break less-cf-:oo'er.: accicent analysis anc operator training anc pro:edures.
H:we ve r, at thi s t ime, we are beginning to look more teeply into aeditional':esign features of B&W plants to consider if any further system modif t:ations are necessary.
The use of once-through-steam-generaters (OTSG) ir. B&W :l ar.as has an opera-tional acvantage in that it provides a small cegree Of steem superheet, as contrasted with the conventional saturatec U-tu:e steam generator.
In accition, it provices for less water inventory thus making a steam line break less severe. Howver, the relatively low water inventory with the presence of a liquid-vapor heat transfer interface in the active heat transfer :ene closely couples the primary system to the steam generator conditions with a consequently high sensitivity to feetwater-flow rate perturb:: ions. Enclosure I to this letter addresses system problems and staff concerns in this area. At present, we are inves:igating whether S&k plants are overlysensitive to feedwater transien;s, cue to the' 0750 con ect, as cou;lec with the pressuri:er si:ing, !;5 :esign, and PCP.V/ reactor tri: se; ::ints.
As ;ar: Of tne ;os: TM:-2'effer:, Oe:aile: ara'.ysts.a.e :ee-.a:e ' uncer.
cc ling transients f or B&k :' ants.
however, :ve
- ,0 :*e ser.sitivity Of the 735 :es gr, S&i :len t have als; been ex:e*1erci.g a 7:e- :# *etatively stsere 0.tr:0 ling eve" 5.
Mj' ch 9 0.12 z w) h 5 i,
2
Mr. M. 3. D arri s-F:r y ur inf rmation, NRC is initiating a resear:n tas. :: :vantitatively assess B&W system designs, inclucing the integrated contaol system, timed at icentifying obvious accident sequences leading to cere damage having a high frequency as compared to the Reactor Safety Stu:y, see Enclosure 2.
(A complete determination of risk will not be attem;ted). The objective of this assessment is to identify high-risk accident se;uences (including TMI implications) utili:ing event tree and simplified f ault tree analyses.
Included will be estimation of release categories, appr:ximate quantifi-cation of expected frecuency cf selected event se:vences and sensitivity studies for reliatility of operater response.
The stu:y will focus on the risk implications of the sensitivity of the E&W :esi 9 and on the i
potential interactions arising.frer the integrate: ::t--f system.
We' esti ate this study to oe complett-: in about six =0*.ns.
We will use the Crystal River, Unn 3 plant as :ne reference: ft:i:f o be analyzed.
We have been holding generic-discussions with Bacco:k an: Wilcox Company
- encerning this matter. However, system sensitivity ;0 fetcwater transients involves balance-of-plant equipment and systems as well as the nu: lear steam supoly system, and such plant spe:ifi: characteristics. 5% be censicered.
We are aly consicering whether it is necessary to hal: :ursions cf the constr,:ction of B&W plants, pending the outcome Of int -t11 ability assess-me n-As a preliminary consideration, we have icentifie: in:se systems and com*;:nents that may be impacted by pessible design :hanges as. result of this study. Enclosure 3 is a preliminary listing of su:n systems and components.
Unde
- the authority of Section 132 of the Atomi: Energy 1:t of 1954, as amen:ed, and Section 50.54(f) of 10 CFR Part 50, a::iti:tal information is requested to allow us to determine whether it is ne:essary to halt all or portions of the construction of your plant pen:ing :ne results o# our st ucy.
We request you provide:
a) Identify the most severe overcooling events (consi:ering both anti-cipated transients and accidents) which could oc:ur at your facility.
These should be the events which causes the greatest inventory shrinkage. Under the guidelines that no cperator action occurs before 10 minutes, and only safety systems can be used to mitigate the event, each licensee should show that the : re remains adequa,tely cooled.
- )
Icen-ify whether action of the ECCS or US (:- ::e-n:r a: ion) is ne:essary :: Orc:e:: the core # liowing : e
- s*. seven over.
- coling transien; icen;ified.
If :hese syste s : e e:vice:,
you should show that its de1ign criterion ':-
- e u :e* Of a: unien :ycles is a: ecus:e, :ensi:ering a-d.C -nes ':r ex:essive :: ling transien:s.
Mr. H. G. Parris c) trovice a schedule Cf completion cf installation of the icentified systems ated corconents.
c) Icentify the feasibility cf halting installation of these systems and components as compared to the feasibility of completi g installatien and then effecting significant changes in these systess and components, e) Co cent en the OTSG sensitivity to feedwater transien:s.
f) Provide recomendations on hardware and procedural changes related to the need for and methods for damping primary system sensitivity, to perturbations in the 0753, Incluoe cetails on any cesign acequacy studies you have done or 'have in progress.
We are sentiin; si-ilar letters to all utilities nc1:ing construction per.its fer :lants wuh S&W nuclear stea su;;'y systers.
We request your reply by December 3,1979. We celieve that a meeting witn you anc the other utilities together with the staf# anc the Sabcock and Wilcox Company to ciscuss this matter wovic be beneficial to all parties.
At that time, we will provide furtner cetails or. tne Crystal D,iver Stucy.
We are scheduling such a meeting for Novem:er 5,1979 at 10:00 a.m. in Room 0-422 at our of fices in Bethesca. 7920 Nor'olk Avenue, Betttica, Maryland.
D1 ease call Dr. Anthony Sournia at (301) 492 7200 if you have.ny euestions concerning this letter.
Sincerely,
/
.u a e
. r~
Harold R. Denica, *)irect or Office of Nuclear S.eector Regulation
Enclosures:
As stated cc:
See next page 4
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Mr. H. G. Pa rris CC*
Herbert 5. Sanger, Jr., Esc.
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General Counsel Tennessee Valley Authority 400 Commerce Avenue E11B33 Knoxville, Tennessee 37902 Mr. E. G. Be e sley Tennessee Valley Autht rity 400 Commerce Avenue, k'0Cl31C Knoxville, Tennessee 37902 Mr. D. Terrill Licensing Engineer Tennessee Valley Aut.he-ity 400 Chestnut Street iorer - !!
Chattanooga, Tennessee 37401 Mr. Dennis Renner Babcock & Wilecx Company P. 0 ' Box 1260 Lynchburg, Virginia 24505 Mr. Robert B. Borsum Babcock & Wilecx Company Suite 420 7735 Old Georgetown Road Bethesca, Paryland 20014 9
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