ML16341D296
| ML16341D296 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 07/08/1985 |
| From: | Dircks W NRC OFFICE OF THE EXECUTIVE DIRECTOR FOR OPERATIONS (EDO) |
| To: | Asselstine NRC COMMISSION (OCM) |
| Shared Package | |
| ML16341D297 | List: |
| References | |
| IEIN-85-45, NUDOCS 8507180084 | |
| Download: ML16341D296 (38) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 JUL 0 8 )885 MEMORANDUM FOR:
Commissioner Asselstine FROM:
SUBJECT:
William J. Dircks Executive Director for Operations DIABLO CANYON - SEISMICALLY INDUCED SYSTEMS INTERACTION Your memorandum of June 17, 1985 requested information on staff actions regarding the subject with respect to IE Information Notice No. 85-45.
The potential for a seismically induced systems interaction (SISI) between the movable in-core flux mapping system and the seal table was identified by PGSE in March 1982 as a result of its SISI program and was reported to the NRC in the Final Report for the program in May 1984 in accordance with the staff approved program.
Modifications were completed for Unit 1 in December 1983 and for Unit 2 in June 1985.
Enclosure 1 is a listing of events and actions, including references to the Diablo Canyon SISI program which had been developed by PGSE in early 1980 in response to a recommendation by the ACRS in late 1979.
The specific modifications are listed in Enclosure 2.
Since the problem had been corrected the staff did not take a specific action on this matter for Diablo Canyon as a result of the Shearon Harris notification in June 1984.
Based on a review of the information for the specific inter-action in the SISI program.Final Report and recent discussions with PGRE, the staff finds the specific actions and modifications appropriate.
The staff intends to audit the records and modifications in the plant for this system interaction.
Regarding a notification by Westinghouse, the Westinghouse Safety Review Committee reviewed this matter on May 29, 1985 and determined that it was a
potential unreviewed safety question for operating plants as defined in 10 CFR 50.59 and a potential significant deficiency for plants under construc-tion as defined in 10 CFR 50.55(e).
Westinghouse notified the utilities by telephone within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and written confirmation was provided within ten days.
In early June 1985, the NRC staff contacted the Regulatory Review Group of the Westinghouse Owners Group regarding additional information which subsequently was provided to the staff in a letter dated June 10, 1985 (Enclosure 3).
8507180084 850708 PDR CONMS NRCC CORRESPONDENCE PDR CONTACT:
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To date the staff has made no determination if Westinghouse has complied with applicable NRC reporting requirements, and further, is determining whether this matter should have been separately reported under 10 CFR 21 by PG&E, apart from its submission of the SISIP Final Report.
The staff plans to assure that all Westinghouse licensees have received the IE Notice and the Westinghouse notification, and that they have made appropriate evaluations.
This will be accomplished by a Temporary Instruction issued to the Regions.
Enclosures:
- 1. Diablo Canyon Units I and 2
SISI History for In-Core Flux Mapping System/Seal Table
- 2. Diablo Canyon Units' and 2
SISI Modification
- 3. Letter from G. T. Goering,
- WOG, to F. Miraglia, NRC, dated June 10, 1985 (with attachment) cc:
Chairman Palladino Commissioner 'Bernthal, Commissioner Zech
'ECY OPE OGC J'.
M. Cutchin (Signed) Vlilliamj.
Dirc'illiam J. Dircks Executive Director for Operations
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To date the staff has made no determination if Westinghouse has complied with applicable NRC reporting requirements.
The staff plans to assure that all Westinghouse licensees have received the IE Notice and the Westinghouse notifi-
- cation, and that they have made appropriate evaluations.
This will be accomplished by a Temporary Instruction issued to the Regions.
Enclosures:
- 1. Diablo Canyon Units 1 and 2
SISI History for In-Core Flux Mapping System/Ical Table
- 2. Diablo Canyon Units lend 2
SISI Modification
- 3. Letter from G. T. Goerin+
- WOG, to F. Miraglia, NRC, dated June 10, 1985 (with attachmen
)
cc:
Chairman Palladino Commissioner Roberts Commissioner Bernthal Commissioner Zech OPE OGC SECY William J. Dircks Executive Director for Operations P
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UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON D C 20555 MEMORANDUM FOR:
Comm'ioner Asselstine FROM:
SUBJECT:
W liam J. Dircks ecutive Director for Operations DIABLO CANYON - SEISMICALLY INDUCED SYSTEMS INTERACTION Your memoran re ardin t
f June 17, 1985 requested information on staff actions b
ct with res ect to IE Information Notice No. 85-45.
0 9
g SU The potent 1 for the movabl in-core PG&E in rch 1982 re orteh to the NR p
seismically induced systems interaction (SISI) between flux mapping system and the seal table was identified by gd dgggp g'
in Ma 1984 in accordance with the staff a
roved ro ram.
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g Modifications were completed for Unit 1 in December 1983 and for Unit 2 in June 1985.
Enclosure 1 is a listing of events and actions, including references to the Diablo Canyon SISI program which had been developed by PG&E in early 1980 in response to a recommendation by the ACRS in late 1979.
The specific modifications are listed in-Enc osure 2.
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Since the problem had been correct the staff dM not ake a specific action on this matter for Diablo Canyon as a
ris notification ld d ~
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program Final Report and r t
iscussions with gG&E, the staff finds the 4
f'".i actions and modifications ypropr ate.
however, (he staff intends to vev~
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audi t, ef-'
L. Ia, records and modifications in the plant)mr Hsds ggIb InIRrtg c,Isggsg.
Regarding a noti ication by Westinghouse, the Westinghouse Safety Review Committee review d this matter on May 29, 1985 and determined that it was a
potentialysn~e4red-safety question for operating plants as defined in 10 CFR 50.59 and a potential significant deficiency for plants under construc-tion as defined in 10 CFR 50.55e.
Westinghouse notified the utilities by telephone within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and written confirmation was provided within ten days.
In early June 1985, the NRC staff contacted the Regulatory Review Group of the Westinghouse Owners Group regarding additional information which subsequently was provided to the staff in a letter dated June 10, 1985 (Enclosure 3).
CONTACT:
H. Schiei ling, NRR X-27100
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To date the staff has made no determi ation if Westinghouse has complied with applicable NRC reporting requirement The staff plans to assure that all Westinghouse licensees have received he IE Notice and the Westinghouse notifi-
- cation, and that they have made appropriate evaluations.
This will be accomplished by a Temporary Instruction issued to the Regions.
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Enclosures:
- 1. Diablo Canyon Units 1 and 2
SISI History for In-Core Flux Mapping System/Seal Table
- 2. Diablo Canyon Units 1 and 2
SISI Modification
- 3. Letter from G. T. Goering,
- WOG, to F. Miraglia, NRC, dated June 10, 1985 (with attachment) cc:
Chairman Palladino Commissioner Roberts Commissioner Bernthal Commissioner Zech OPE OGC SECY William J. Dircks Executive Director for Operations
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Enclosure 1
Diablo Canyon Units 1 and 2
In-Core Flux Mapping System / Seal Table Summary of SISI Activities November 1979 Mid 1980 October 1980 March 1982 September 1982 November 1982 June 1983 July 1983 October 1983 December 1983 February 1984 April 1984 May 1984 October 1984 ACRS recomnends that PG&E perform a seismically induced systems interaction program (SISIP)
PG&E develops SISIP for Diablo Canyon Units 1 and 2
(submitttal of a final program in August 1980)
In SSER-ll NRC staff reports on initial SISIP results and finds program acceptable.
DuriAg its SISIP PG&E identifies the potential for a specific SISI between the movable in-core flux mapping system as the interaction source and the seal table as the interaction target.
PG&E request Westinghouse to review and analyze the potential SISI.
Westinghouse provides PG&E with information and recommended modifications.
PG&E issues a Design Change Notice (DCN) for Unit 2 modifications.
PG&E issues a Design Change Notice (DCN) for Unit 1
modifications.
PG&E submits Status Report on SISIP for containment; no reference is made to the specific SISI.
PG&E completes specific SISI modifications for Unit l.
PG&E completes specific SISI verification walk-down for Unit l.
NRC inspects documentation and implementation of selected SISI's for Unit 1
( Inspection Report 50-275/84-09);
the specific SISI is not included.
PG&E submits SISIP Final Report, including the specific SISI for Unit 1 (Section 9.3.1.1 and Attachment 7-8).
PG&E submits Unit 1 interaction documentation sheets for SISIP Final Report; includes the specific SISI.
J
Enclosure 1
January 1985 April 1985 April 1985 June 1985 NRC inspects documentation and implementation of selected SISI's for Unit 2 (Inspection Report 50-323/85-05);
the specific SISI is not included.
PGKE submits Final SISIP Report, Rev.
1, for Units 1
and 2; includes the specific SISI for Unit 1 (Section 9.3.1.3 and Attachment 7-B) and for Unit 2 (Section 9.4.1.3 and Attachment 15-B); resubmitted in May 1985.
NRC issues SSER 31 which includes its SISIP evaluation (Section 4.2); it does not address the specific SISI.
PG&E completes modifications and verification walk-down for Unit 2, including specific SISI.
Enclosure 2
Diablo Canyon Units 1 and 2
SISI Modifications 1.
Welded fixed frame base plates to trolley beam.
2.
Replaced 9.375 inch cap screws with ASTM A325 bolts (or equivalent) of same size.
3.
Modified and installed movable frame anchor angles.
4.
Installed seismic anchor bracket.
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1985 Hr. Frank Hiraglia, Oeputy Director D1vision of Licensing Office of ltuclear Reactor Regulation U.S, Nuclear. Regulatory Coemlission ihshtngton, DC 20555 Sub)ect:
Potential Seisaic Interaction Associated arith the Flux NappIng System in Nestinghouse Plants
Dear Hr. NIraglia:
As you are avare, westinghouse ident1fied a potential'eportable Itew to utilities ~1th Hestinghouse supplied VASSS concerning a possible interaction betxeen the non-nuclear safety flux sapping system and the Class l pressure boundary portions of the seal tablelbottce munted instrumentation (NI)
'uring a postulated seismic event.
This itea was'evieeed by the internal westinghouse Safety Revie~ Coneittee (SRC) on 29, 19B5.
As a result. of th1s meeting, the CNIttee deterained that the ea was a potential unresolved safety issue as defined in 10CFR50.59 for operating plants and a potential significant'eficiency as defined in 10CFR50.55(e) for plants under construction.
En accordance with Westinghouse internal procedures, tIIIties vere tlfied b t le e within twenty-four (24) hours of the dec s on and m tten irwat on was provided'vithln ten (10) days, A representative letter se to ut l'Ities by westinghouse is attached for one It shou)d be noted that the review perforaed by the westinghouse SRC focused on the potential fai'lure of the flux wapping systew due to a postulated seiswlc event and iti potential interact)on with the seal table/BNI.
The consequences of the flux Napping systea, or parts thereof, falling on the seal table/86 mare not analyzed.
It is not certain that the pressure toundary at the interface of the thiable and the thiable guide tube meld be breached.
This is due to the aany flux sapping detector guide tubes, which are not part of the pressure boundary, that are in place batmen the flux sapping system mvab)e fraee and the thiwbles and thiable guide tubes which are part of the pressure
- boundary, Because of the 'large nuaber of different flux aa In s stoa d
5 ns, aounting provisions.
the different esigne s rue ural es gn arrangements, and
'various ways that systee interaction Nay occur in Individual plants, westinghouse does not have tuff I nt Inforaation to evaluate.the degree of protection af orded o aa n ain the integrity of the guide tubes at the seal tab1e for s cific lants.
- Heaver, Nr. Frank HIraglia of the NC contacted the Regu atory ev ev Group
<RRG) of the Hestinghouse Orner's Group requestIng addi tiona1 inforaat Ion.
)
JM 1i '85 B7:52 RED 78i HJK 412-256W743 P.83 OG-151 June 10, 1985 Subsequent to thfs request, several telecons vere held durfng the seek af June 3,
1985 aug the KRC staff, Mestfnghouse and RRG representatfves.
Ourlng these telecons, several fteis ware dfscussed and the Staff requested further RRQ revfev of this ftem.
The lnfor<<ation contained herefn fs provfded fn response ta this request.
The fnteractfan of the flux <<appfng system fifth the seal table/8HI resu}ting 1 n a breach of the RCS pressure boundary ls cons dered to be a fee rababf lfty
~ve The basfs for thfs conclusion fs that for s
n eract on to occur, a
)or portion ot the flux <<applng system Iust be located above the sea1 table.
Gfven thfs layout conffguratfon, a potentfal
'fnteractfan could only result from a sefs<<fc event.
Ourfng the occurrence of the seisnfc event, the flux <<appfng systee wwld have to fall fn the structural
<<ode descrfbed below to rupture any guide tubes. If the gufde tubes dfd rupture fn the awfe
. descrfbed beler, a saa11 loss of coo'lant accident veld result.
As also sumerfzed beta',
such an event veld not result fn any care uncovery.
- Thus, the potential for these events. to occur and result fn an unacceptable condftfon fs considered to be of lcm probabf lfty.
Iased on fnput free NX
. utflftfes lt fs fmportant to note that w e a number of la n t have thfs potentfal ra41oa e ta the physfcal focatfon af the flux aappfng
~qufpeen or e p ants'esf n revfevs a
e a resse t e sefssfc restraint of the flux <<appfng system.
STRUCTURAL KVhLUATION h revfe~ of cee representatfve plant flux <<appfng system was conducted.
'Sfnce all Nestfnpouse plants have dffferent desfgned flux <<appfng syste<<equfpaent conffguratfons.
a representative system eras randomly chosen from that group of plant conffguratfons vhfch have thefr flux <<a ln transfer cart hun fram a rail car and track over the s
1 table rhfch is the t case.
T e results s rev ew n
ca e t at there ex sts a po entfal for e
allure of nat
<<are than 3 BH7 guide tube stubs.that protrude above the seal table, lf the transfo fng a sefsafc event and fall onto the seal table, For thfs erst case scenarfo to occur, the follcwfng issuaptfans
~ere Nade:
A sefsafc event ls occurrfng The transfer cart ls not equfpped fifth lever lateral restrafnts and the cart fs allmed ta svfng durfng the sefsafc event.
The rail car ls nat equipped vfth wheel restrafnts to prevent the car fraa )u<<pfng off track and fallfng.
(h case fifth wheel restraints sas also lnvestfgated).
The lateral translation of the svfngfng <<ation fs large enough that tm (2) af the support Ieebers (threaded rods and turnbuckles), Qfch connect the transfer cart framfng to the above rail car, 'fall due to hfgh bendfng stresses.
The transfer cart draps vertically vhf le stf ll <<avfng laterally and scee part of its frawfng fapacts agafnst the s
s of 3 or less gufde tube stubs sfaultaneausly.
(The rans er car as onl enau h energ ta faf 1 no core than 3
tubs).
No credf s taken for the lnterconnectfng effects of the fluk thfmble tubes or transfer tubing to deepen the fmpact foadfng on the gufde tube stubs or to absorb the kfnetfc energy fn the swfngfng transfer cert.
The rafl beaas abave the transfer cart do not fall.
6662g:12
L I
. 'lN.11 '85,87: 55 AD 78f BLDG 412-PS&6743 P.84 OQ-l51 June 10, )985 Yho fixed frame that supports the transfer devices/drive
<<echanisms do not fal),
A previous Westinghouse ana)ysis on this equlpeent plants has shovn this to be so.
Once the transfer cart hits the sea) table or guide tube stubs, tho action of the cart ls stopped.
Tho fo))ovlng addltiona) infor<<ation shou)d be nofod:
Ko failure to the seal table itself or to the guide tubes below the seal table cou)d be postulated.
If the transfer cart were to drop vertically, with no lateral trans)ation, no guide tube stubs ri) 1 fal).
If the transfer cart is equipped with )over latera) restraints and rheo) restraints it cannot fail or drop onto the seal tab)o, These )oror restraints are typically located on the cart frailng at the e)ovation of the thimb)e tube isolation valves.
If the transfer cart were to fa/l and hit the thieb)e tubing, the tubfng rill fall but no breach of the pressure boundary vou)d occur.
If the transfer cart fraeing i<<pacts 4 or mre guide tube stubs sieu)taneous)y, none of the impacted stubs rill fa!).
The swage lock coapresslon fittings at the top cf the guide tube stubs are stronger than the guido tube stubs.
Fal)ure of the flux mapping system, causing it to fa) l ril) occur with or rithout rhoe) restraints if no )ower )atora) supports are provided.
LOCA EVALUATION The ee)tip)e failure of a nuaber of gu'lde tubes resu)ting froa the failure of the flux aapplng system vou)d be <<ore severe than the failure of the single guide tube vhlch is bounded by sea))
break LOCA analyses doculented ln Chapter
)5 of the FSAR.
The consequences of the av)tip)e fal)ure of guide tubes ls, however, not intolerable.
The safety in)ection systol on each plant design vill be capab1e of <<ltlgating the consequences of the <<u)tip)e fai)ure of a nuabor of guide tubes.
Due to the )ocatlon of this 'sea)1 break LOCh,'the nu<<ber of guide tube failures which can be tolerated is
) lalted.
S<<a) 1 break LOCA ana)yses for various aumbors of instruwent tube fal)ures are not avai)able.
However, the attachment provides an eva)uation of tho ccesequences of au)tip)e guide tube failures and presents a conservative estl<<ate of the nuebor of guide tube failures which cou)d be tolerated under the current sma))
break LOCA evaluation aode) licensing llelts vlthout exceeding tho lielts i<<posed in appendix K end
)OCFR50.46.
'VALUATION OF POSSISLE STRUCTURAL MODIFICATIONS As noted ln the Nastlnghouso coeeunlcations vlth utilities, lt vas recoeaendod that the flux Napping system be restrained to prec)ude any possible interaction with the seal tab)e/NI, westinghouse has perforlod a seismic analysis of two typical flux upping systo<<s
<R6> and identified posslbl'e syste<<<<odl flcatlons. 'odifications to the <<ountlng configurations vere required to insure that the flux aapplng systee rou)d rithstand seismic loads without structura) co))apses In ono c vere r
at the rhoe) loc n
ae of the structure a
the lsolati
)vo&
p)etfor<< (at the seal table>.
A second case requ red ex s
ng whee) ply I,b4dtd t
Rt t It tkh I tt f
the structure.
6662g:12
1 1
JlÃ.5% '8S, 87:59 FRD 781 KEG 442-256W743 OG-iSi June
)0, 1gsg The wheel restraints provide a posftfve connectfon between the loveable frame and the structure which supports the Reel rails, The wheel restraints are intended to prevent the eweable frame frea rollinO alceg the rails and to prevent Keel liftoffso that the structure does not fa.ll off the rails.
The restraints at the fsolatfon valve platforl (bottos of the structure) provide a posftfve connection to the surrounding support structure.
These restyafnts are intended to prevent large seismic dfspface<<ents at the bottce of the structure, These large dfsplacements cause large loads which say overstress the vertfcal'tructural <<e<<hers and <<ay cause structural collapse.
. Kfth the above <<entloned eodiffcatfons, stresses fn the structural eelbers are held efthfn acceptable ll<<its during a postulated seismic event.
In siary, tbe RAG has revfeved thfs f tea ln light of the fnforaatfon contained herein and has concluded that there fs no significant safety fssue requiring fllsedfate Lfcensee action.
%eever, there is a need to revfee the 1>> out n
u the flux aapping systea to preclude any s
le interaction vfth the seal table/BHI.
This act d
on a an fl fnlt ated Were ed.
hny require
<<odfffcatfons h
be <<a 1
ent with 4
spec c p an ou age and construction schedules.
SSOORY o
This potential problem could result if the design of the flux Napping equfp<<ent eountfng dfd not address the possfb1lfty of'efs<<fc interact/on fifth the instrueent.tube sea) table.
o A.revfer of the e)rst case interaction event does not result ln unacceptable consequences.
o eased on utility input to the RRG va believe that a number of plants do not have a probfea due to equipment location or by having eddreseed the sefsefc desfgn of the flux aappfng equfpient durfng plant design.
Note that thfs fs the case vfth Diablo Canyon and Shearon Harris has previously identified this issue to the Staff.
The need for the sefs<<fc desfgn vas ldentlfled ln plant spccfffc design revfews for the above
<<entfonef plants.
o As a result of the Hestfnghouse letter to the Utilities individual plant reviews are taking place to identify any potential problees.
. The 55 has reviewed this ftem fn lf ht of the fnforiatfon contained herein and has concluded that there is t safet 1
u re frlng d1ate Lfcensee action.
Hcwever, there ls a need to review the layout and the es gn o the flux <<applng systee to preclude any possible fnteractlon vf.th the seal table/BHI.
lfe believe this action vile be perforaed on a plant specfffc basis by lndfvidual utilities and correctfve action tnftfated ~her'e required.
hny required <<odfffcatfon to the f)ux aapptng system +@lid be made consistent fifth specific plant outage and construction schedules. '@g C.c.
- Yery tru'ly yours, EC4~
6662gr12
- 6. T. Goerfng, Chair<<an, RAG Subcoeefttee Hestfnghouse Seers Group
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6ener1c Letter to Ut111t1es re: Potent1al for System Interat1on Between the Flux Mapp1ng System and the Seal Table Rcspons1ble Ut111ty D ar Potent1al for System Enteract1on Between the Flux Mapp1ng System and the Seal Table Due to Se1sm1c Loads Oq May d)9, 1985, Mest1nghouse conducted an 1nternal Safety Rev1ew Conelttee
~et1ng to'eview the potent1al for system 1nteract1on between the flux sapp1ng system and the seal table/bottom mounted 1nstrumentat1on (BMI) rqsult1ng from a postulated se1sm1c event.
Th1s equ1pment 1s suppl1ed at n)clear power fac111t1es us1ng Mest1nghouse suppl1ed Nuclear Steam Supply S stems.
ntr 1onP
')e cem1ttee found th1s 1tem to be a potent1al s1gn1f1cant def1c1ency as dtf)ned )n )OCFR50.55(e).
'r er n
n s l
Qe coae1ttee found th1s 1tem to be a potent1al unrev1ewed safety quest1on as d f1ned 1n 10CFR50.59.
I accordance w1th our report1ng procedures, Ut c
. vjs not1f1ed v1a a telecon on 30 Nay 1985.
Th1s letter prov1des wr1tten c nf1reat1on of the potent1al 1tem discussed 1n that telecon.
T)1s 1ssue of poss1ble 1nteract1on between the non-nuclear safety flux mapp1ng system and the Class l pressure boundary port1ons of the seal table/BMI due to a,se1sm1c event has been 1dent1f1ed by&a, o se 1n lants lo rr1s.
Investlgat1ons performed for these plants we e
l m1ted to the structural capab111ty of the flux mapp1ng system under se1sm1c 1 ads.
The 1nvest1gat1ons sho~ed that w1thout adequate restra1nts the flux pp1ng systems could collapse and poss1bly fall on the seal table/BMI.
l e consequences of the flux aepp1ng
- system, or parts thereof, fill1ng on the s al table/BMI were not analyzed.
It 1s not certa1n that the pressure b undary at the 1nterface of the th1mble aqnd the th1mble gu1de tube would be b,cached.
Th1s 1s due to the aeny flux mapp1ng detector guide tubes.
wh1ch afe not part of the pressure boundary, that are 1n place between the flux
'kpp1ng system movable frame and the th1mbles and th1mble gu1de tubes wh1ch a'e part of the pressure boundary.
0 86n/JN/5-85
a t
'F ROH (HOH)86'8'85 16I 47 HOe 12 P4GE 2
auld requ1re the equ1valent of the complete severance of one flux th1mble gu1de tube at the seal table to const1tute a small loss of coolant acc1dent (loca).
I s
should be noted that only two of the many des1gns of flux Napp1ng systems re 1nvestlgated.
Parts of these systems mre located d1rectly above the al table.
Among the var1ous
- systems, some have no bove the eal
- ble, some o al bove the seal i ble 1le others as n the case of~ ~
ose 1nvest1 ated omevhere between these two con t1ons.
8 cause of the large number of d1fferent flux mapp1ng systems
- designs, unt1ng prov1s1ons, the d1fferent BOP des1gned structural des1gn a rangements, and var1ous says that system 1nteract1on may occur 1n 1nd1v1dual p ants, ve do not have sufflc1ent 1nformat1on to evaluate the degree of p otect1on afforded to ma1ntaln the 1ntegr1ty of the gu1de tubes at the seal table for your plant.
Therefore, 1t 1s recommended at ut111t v
tlgate t
the restra1nt s stem under s 1sts~loads, you have any quest1ons or requ1re any add1t1onal ass1stance, please contact e lt Pro)ect Off1ce.
0 86n/33K/5-85
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SED-SA-01030 (HOH)86 ~ 18 ~
85 16<47 HOo 12 PRGE June 7, 1985 hTTACHMENT CONSZ{}UENCE OF h SMALL BREAK LOCA DUE TO MULTIPLE ZNSTRUMENT TUBE FAZKURES The breach af the reactor coolant system (RGS) px bounda at libration belov the steam generatax secondary side conditions<
term cooling af tho RCS is not impaired.
The reactor coolant, em vill simply depressuriae to the point, vhexe the decay heat and ed safety infection flov induced volume ave11 oyxels the volume nkage from <<ass last thraugh the bxcak and decay hea% removal ta steam generator secandary sides.
Igain no core mcovery vill lt and no cladding heatup villoccur.
ion ays
. the
.rosu
." Hce px'e aec ver, if the number af <<ultiplo tube failures xesult>> in a sure equilibration point vhich is belov the steam generator ndary side Conditions, the RCS vill depressuriee to a pressure essure the.'ottom af the xoactar vessel due to the failure af an instrument tube villresult in a small break lass of coolant accident (LOCA).
Tho, failure af a single instrument tube due to the failure of a veld fs considered and is bounded by the small break LOCA analyses found in chapter 15 of each utilities FSAR.
Since the <<ass flov rate through the break defined by the tube failure vill exceed the chaging capability cf <<oat plants, the failure of a single instrument tube villtypically result in the deprossurleatian.of the RCS causing a reactor trip due to lov pressuriLer pressure.
The con'tinued depressurisation vill generate an 6-signal and pumped safety in)ection iiov villbe started.
Tho roactar coolant system Oil). continue ta depressurise ta the paint vhere tho pumped safety infection <<ass flov rate ovals th>> <<ass flav rate through the bre'.
This pressure point is defined by the intersection af the pum ed safety in)ection <<ass flov versus pressure characteristic
, e and the critical <<ass flov versus pressure curve for the given break area as discussed in reference 1.
The RCS <<ass inventary vi11
<<~alite at this pressure and lang tera caro cooling <<ethods can be iapLi<<ented.
No core uncavery villresult in this scenario and
.conliequently there villbe no cladding heatup.
Tho, <<ultiplo failure af a number af instrument tubes increases the
<<as'nventary lost thorugh the break corresponding ta the increased bronc flov area defined by the <<ultiple tube failures.
Multiple faiguro af a number af instrument tupes villresult in th>>
dephossurisatian af the RCS resulting fn I reactox txip du>> to los rossuriaer pressure.
Continued depressuriaation villagain result
~
6-signal being generated and pumped safety in)ection flov vill be tarted.
Tho b eak <<ass flay rate due to <<ulti lo nstrumont tube fai, ures vil) also m
al lov considexat ans and a new ressnre e
ret on yoint will he OetfneC. PzovBR ~~t the n
er o
<<u p e t a
urea s no result in a pressure olpl
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C1IOH)86o 18m '5 16 '9 HO ~ 12 PAGE
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~ 'D-SA-01030 3 0 June I, 1985 Saloudek critical <<ass flov correlation for subcooled Xiquid flov at an enthalpy of 550 STU/1bm.
3.
Minimum Safety in)ection flov for all Sestinghouso designed I
WRs t 1300' a
The results of this. calculation indicate that the <<ulti le failure of struuent tubes sill result in a pressure equilibration paint e 1300 psia for a plant vith the smallest amount, ef safety ction flov.
4 iy abo inj
~
e ~
abo. e the steam generator secondary side pressure.
The reactor coolant system vill again eyxilibrate at a pressure where the decay heat and pumped safety Infection flov induced volume sveil equals the volume shrinkage from mass lost through the break and heat removal to the'team generator secondary sidese Tho RCS pressure vill remain relatively constant at these conditions unless some action is taken.
At this pressure(
in this situation, the mass flow lost through the hre'ak exceeds the <<ass flov replacement from the pumped safety in)ection flov and the RCS mass inventory. continuously decreases to
.. the'oint that Core uncovery vill result.
Bince the pressure ho&dary breach is at the bottom of the reactor vessel, there is no
<<echanism which villpermit the break to remove more of the decay heat and pumped safety infection induced fluid volume avail.
Consequently(
the limits of 10CPR50.46 and Appendix K may be exc'ceded.
The, narinuu nuuber cf'instruuent tube !ailures nay be ca1culated by 4etermining the number cf tube failures which vill result in a 4eptessuritation to an equilibration above the steam generator haec'pndary side conditions.
The <<inimum flov area for the failure of
- an instrument tube is defined by the thimble plugs in the instrument tube.
This flov area i¹ Oe0002i053 ft~.
This fkov area will 4ef ne a family of <<ass flov versus pressure curves for vaious ers of failed instrument tubes assuming some critical <<ass flow co elation.
Knowing the plants characteristic
<<inimum safety infection <<ass flov rate versus pressure curve villdefine a number oi gntegsection points at various pressures.
The last intersection point which is above the expected steam generator secondary side con itions vi11 define the <<aximum number of tolerable instrument t
failures.
1 unding calculation of the number of acceptable instrument tube
'ai urea vas performed vith the following conservative assumptions!
1<
Minimum aooeptable RCS e~ilibration pressure of 1300 psiae I'.
.'. Rei rence 1 ! NCAp-9600( >Report on Small Break Aocidents l'or lfestinghouse NSS Bystemt1, June l979.
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(~~ QD/Q EDO PRlNCIPAL CORRESPONDENCE CONTROL
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