ML19199A387
| ML19199A387 | |
| Person / Time | |
|---|---|
| Site: | Crane  |
| Issue date: | 12/21/1972 |
| From: | Anthony Giambusso US ATOMIC ENERGY COMMISSION (AEC) |
| To: | John Miller METROPOLITAN EDISON CO. |
| Shared Package | |
| ML19199A389 | List: |
| References | |
| NUDOCS 7905020009 | |
| Download: ML19199A387 (13) | |
Text
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Occket ;,c. 50- 320 EC 21 I'etropolitan F;iiscn Ccnpany
.':r. Jcnn G. n ' l er ra:
Vice ?Meident
?, C. Eox 542 F.erdinc, Fenraylvania iggc3 Centle :en:
Se ?.:e.ularcr/ staff's centirmirc review of reacter pcwer plant safety ircicates tnat the ocnsequences of pcstulated pipe failures outsica of the ccntairnr.nt structure, including the rapture of a nain steau or feedwater line, need to be adequately doctn.cnted and analymd by licensees and applicants, ar4 cvaluated by the staff as seen as possible.
Critericn 1.o. 4 of the C.... iasion's Gereral Design Criteria, listed in Apperc12 A of 10 CJF. 50 requins tnat:
" Structures, systecs, ard ccapenents igcrtant to safety shall be desirned to acccer.cdate tim effects of cr4 A be ec patible with the emircrxcntal ccrcitiens associated with e 1 cperation, raintenance, testing ar4 postulated accidents, incluMy Icsswf-ecclant accidents. 2ece structures, systecs, and ccgr.ents shall te appmpriately protected against dyrs.ic effects, including the effects of H w ies, pipe whippirg, and discim%ne; fluics, that =ay result f x. equiprent failures and frus events and corditicre cutside tac nuclear pwer unit."
2.e previcus version of the Cn % sien's General "esign Criteria also reflects the above requir ments.
"Iras, a raclear plant should be designed so that the reactor can be scat-dc.'n and caintaired in a safe shutdcwn cordition in the event of a postulatec rapture, cutside ccn'ainment, of a pipe ccntaining a hi;-;n energy fluid, inclnHng the double ended rapture of the largest pipe in the rain steam ard feedwater syste rs.
Plant structures, systers, arc ecrperents irecrtant to safety shcald be designed and locatec in the facility to accccrriate the effects of sucn a pcstulated pipe failure to the extent necessary to assure that a safe shutdcwn ccrditicn of the reactcr ca. i;e acecx;rlished and *ntained.
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"etrocolitan Edison Cc:cany _.
as Based on the linitad information in the Three "fle Island 2 PSAR, we are unable to determine whether or not the plant is adeouataly protected from the consecuences of postulated stean or feecwater cine failures outside of the containment structure.
Discussions with representatives of your company in an 3ttenet to better understand tne desicn and layout of the steam and feedwater systems have not been at all helpful in this regard.
We request that you provide us with analyses and other relevant information needed to determine the consequences of suah an event, using the guidance provided in the enclosed general information request.
The enclosure represents our basic information requirements for plants now being con-structed or coerating. You should determine the applicability, for the Three Mile Island Unit 2 facility, of the items listed in the enclosure.
If the results of ycur analyses indicate that changes in the design of structures, systens, or components are necessary to assure safe reactor shutdown in the event this postulated accident situation should cccur, please provide informacion on your plans to revise the design of your facility to accommodate the postulated failures described above. Any design modifications procosed should include anoropriate consideration of the guidelines and requests for information in the enclosure.
We will also need, as soon as possible, estimates of the schedule for design, fabrication, and installation of any modifications found to be necessary.
Please inform us within seven (7) days after receint of this letter when we may excect to receive an amendment with your analysis of this postulated accident situation for the Three Mile Island Unit 2 facility, a description of any proposed modifications, and the schedule estimates described above.
In addition, we require a summary description of the steam and feedwater system layout vis-a-vis other critical plant systeas which should be included in this seven day resconse.
Sixty copies of the amendment should be provided.
A copy of the Commission's press announcement on this matter is also enclosed for your information.
Sincerely, Criginal s;7,4 _,.
y A. Giambusso, Decuty Director '
for Reactor Projects Directorate of Licensing
Enclosures:
As stated cc:
See next page
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Met rpclitan '-ciscn Ccts:any Eased cn tne linited infomaticn we presently have available to us en Inree Mile Island Unit 2, we are unable to concluce that the plant is adequately p ttected fran the ccnsequences of pcstulated stea:s cr feed-water pipe fM 7 ures cutside of the containnent structure.
'le request that ycu provide us with arJdyses and other relevant infomaticn needed to dete m ine the ccnsequences of such an event, using the pidance provided in the enclcsed general infomaticn request. Tne enclosure represents our tssic infomaticn requiremnts fcr plants ncw being ccn-structed cr cperating. You shculd deter-ine the applicability, fcr the Tnme Mile Island Unit 2 fan lity, of the item listed in the encicsure.
d If the msults of ycur analyses indi:ste that charges in the desi;;n of ssructures, syste.m, cr ccnpcnents are necessa / to assure safe reactcr shutdcwn in tne event this postulated accident situation shculd occur, please f
prc7 ce inforaticn en ycur plans to revise the design of your facilit:/ to acccrrecate the pcstulated failures described abcVe. Any design :mdifi-caticca prcpeced shculd include appropriate censideraticn cf the guidelines and requests for infomation in the enclosure.
Ele will also need, as scen as pcssible, estinates of the schedule for desip, fabricatien, and installation of ari/ c:cdificaticns found to be necessa' J.
Please infcm us within 7 days after receipt of this lester unen we ray expect to receive an amndnent with your analysis of this pcstulated accident situation for the Three Mile Island Unit 2 facDity, a descripticn of av prcpesed rcdifications, and the schedule esti ates descritec abova.
Sixt'i ccpies of the amncant shculd be provided.
A ccpy of the Ccnnissicn'a press anncuncerent cn this nntrer is also enclcsed for your infematicn.
- Sincemly, A. Giantusso, Ceputy Directer for F.eactor Projects Directorate of Licensdr4;
Enclosures:
As stated cc: See next page h
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e Fonn AIC.318.Rev. M3) AEOf 024 r u s m e ewest ee % rv.
- = r-e 1972 466 993
O General Information Required for Considera*. ton of the Ef fects of a Picin_2 System Break Outside Containment The fo llowin g 14 a general lis t of inf ormation required for AEC review of the e f fects of a piping sys tem break outside containment, including the double ended rupture of the larges t pipe in the main steam and feed-wate r sys tema, and for AEC review of any proposed design changes that may be found necessary.
Since piping layouts are subs tan tially different f rom plant to plant, applicants and licensees should determine on an individual plant b as is the applicability of each of the following items f or inclusion in their submittals.
1, lhe sys tems (or portiens of sys tems) for which protection agains t pipe whip is required should ce identified.
Drotection from pipe whip need not he provided if any of the following conditions will exis t:
(a)
Both of the f ollowing piping system conditions are met:
(1) the service temperature is les s th an 200
- F ; and (2) the design pressure is 275 psig or less; or (b)
The piping is physically separated (or isolated) from structures, s ys tems, or componen ts important to safety by protective barriers,
or res trained f rom whipping by plant design features, such as c on c re te en c as emen t ; or (c)
Following a single break, the unres trained pipe movement of either end of the ruptured pipe in any possible direction about a plas tic hinge fo rme d a t the nearest pipe whip res train t c anno t impact any structure, sys tem, or co=ponent imp ort an t to safety; or M
(d)
The internal energy level associated with the whipping pipe can be da=enstrated to be insufficient to 1= pair the safety function of any structure, system, or ce=ponent to an unacceptable level.
2.
The criteria used to deter =ine the design basis piping break loc sciens in the piping syste=s should be equivalent to the folleving:
(a)
ASME Section III Code Claqs I piping breaks should be postulated to occur at the following locations in each piping run or branch run:
(1) the ter=inal ends; (2) any inter =ediate locations between ter=inal ends where the pri=ary plus secondary stress intensities S (circum-ferential or longitudinal) derived on an elastica 11y 1The internal fluid energy level associated with the pipe breck reaction may take into account any line restrictions (e.g., flow li=1ter) between the pressure source and break location, and the effects of either single-ended or double-ended flow conditions, as applicable.
The energy level in a whipping pipe =ay be considered as insufficient to rupture an i=pacted pipe of equal or greater no=inal pipe size and equal or heavier wall thickness.
,'?iping is a pressure retaining co=ponent consisting of straight or curved pipe and pipe fittings (e.g., elbows, tees, and reducers).
3A piping run interconnects co=ponents such as pressure vessels, pu=ps, and rigidly fixed valves that =ay act to restrain pipe =ovement beycnd that required for design ther=al displace =ent.
A branch run differs fre= a piping run only in that it originates at a piping intersecticn, as a branch of the =ain pipe run.
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calculated basis under the loadings associated with one -
half safe shutdown earthquake and operational plant conditions exceeds 2.0 S_
for ferritic steel, and 2.4 S for austanitic steel;
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(3) any inter =ediate locations between ter=inal ends where the ce=ulative usage factor (U) derived from the piping f atigue analysis and based en all nor=al, upset, and testing plant conditions exceeds 0.1; and (4) at inter =ediate locations in addition to those determined by (1) and (2) above, selected on a reasonable basis as necessary to provide protection.
As a =ini=us, there should be two inter =ediate locations for each piping run or branch run.
(b)
ASME Section III Code Class 2 and 3 piping breaks should be postulated to occur at the following locations in each piping run or branch run:
(1) the ter=inal ends; 4Operational plant conditions include nor=al reactor operation, upset conditions (e.g., anticipated operational occurrences) and testing conditions.
5 S
is the design stress intensity as specified in Section III of the ASMZ Boiler and Pressure Vessel Code, " Nuclear Plant Co=ponents."
6U is the cumulative usage factor as specified in Section III of the ASMZ Boiler and Pressure Vessel Code, " Nuclear Fewer Plant Cc=penents."
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(2) any inter =adiate locations between terminal ence where either the circu=ferential or longitudinal -stresacs derivec on an elastica 11y calculated basis under the 1sadings associated with seismic events and cperational plant conditions exceed 0.9 (Sh*S) c the expansion stresses A
exceed 0.8 Sg, and (3) intermediate locations in addition to these determined by (2) above, selected on reasonable basis as necessary to provide protection.
As a =1nimum, there shoold be two intermediate locations for each piping run or branch run.
3.
The criteria used to deter =ine the pipe break orientation at the break locations as specified under 2 above should be equivalent to the following:
(a) Longitudinal breaks in piping runs and branch runs, 4 inches nominal pipe size and larger, and/or 7S is the stress calculated by the rules of NC-3600 and RD-3600 for h
Class 2 and 3 co=penents, respectively, of the ASME Code Section III Winter 1972 Addenda.
S is the allevable stress range for expansion stress calculated by the rules of NC-3600 of the ASME Code,Section III, or the USA Standard Code for Pressure Piping, ANSI 331.1.0-1967.
8Longitudinal breaks are parallel to the pipe axis and oriented at any point around the pipe circu=ference.
The break area ia equal to the effective cross-sectional flew ares upstrea= of the break location.
Dyna =ic forces resulting from such breaks are assu=ed to cause lateral pipe =ovements in the direction nor=al to the pipe. axis.
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4 (b)
Circu=ferential' creak 3 in i;ing c_as anc 'aranch runs exceeding c
1 inch nominal pipe size.
4.
A su==ary should be provided of the dyna =ic analyses applicaol, to the design of Category 1 piping and ia ac o ed iup;r ta which determine the resulting loacings as a resu : of a a:ed pipe break including:
(a) The locations and nu=ber of cesign ousis breaks on which the dyna =ic analyoes are boa 2c.
(b)
The postulated rupture crientation, such as a circu=ferential and/or longitudinal break (s), for each postulated design basis break location.
(c)
A description of the forcing runctions used for the pipe whip dyna =le analyses including the direction, rise ti=e,
=a gn i t ud e,
duration an initial cond :lons tna: dequately represent the stream vna=ics anc the syste= pre <:aure difference.
jet (d)
Diagra=3 c f =a the=a t cc. =c de. s esec :or the dyna =ic analysis.
(e)
A su==ary of tne analyaes,thich de=enstrates that unrestrained
=ction o f rupturec lines stit net da= age to an unacceptable degree, structure, syste=s, or cc=ponen ts i=portant to safety, such as the c;ntrol roc =.
Circu=feren tal breuts c.re perpendicu'ar tc :ne pipe axis, and the break 9
area is equivalent to the incernal cross-sectional area of the ruptured Dynamic forces resulting f r:= such breaks are assu=ed to separate pipe.
the piping axially, and cause whipping in any cirectica nor=ci to the pipe axis.
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A description should be provided of the =easures,.is applic.able, to protect against pipe whip, clowdown jet and reactive forces including:
(a)
Pipe restraint design to prevent pipe whip i= pact; (b)
Protective provisions for structures, syste=s, and ce=ponents required for safety against pipe whip anc blcudevn jet and reactive forces; (c)
Separation of redundant features; (d)
Prceisions to separate pnysically piping and other ec=penents of redundant features; and (e)
A description of the typical pipe whip rectraints and a se==ary of number and location of all restraints in each syste=.
6.
The procedures that will be used to evaluate the structural adequacy of Category I structures and to design new seis=le Category I structures should be provided including:
(a) The method of evaluating stresses, e.g.,
the working stress method and/o r the ulti= ate strength =etnod that will be used; (b)
The allowable desian stresses and/or strains; and (c)
The load factors and the load ce=binationa.
7.
The design loacs, including the pressure and te=perature transients, the dead, live and equip =ent loads; and the pipe and equip =ent static, thsr=al, and d na=ic reactions should be provided.
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Seis=1c Category I structural elements such as floors, interior walls, exterior walls, building penetrations and the buildings as a whole should be analy:ed for eventual reversal of loacs due to the postulated accident.
9.
If new openings are to be provided in existing structures, the capabilities of the =odified structures to carry the design loads should be de=cnetrated.
10.
Verification that failure of any structure, including nonseis=le Category I structures, caused by the accident, will not cause failure of any other structure in a =anner to adversely affect:
(a)
Mitigation of the consequences of the accidents; and (b)
Capability to bring the unit (s) to a cold shutdown condition.
11.
Verification that rupture of a pipe carrying high energy fluid will not directly or indirectly result in (a)
Loss of redundancy in any portion of the protection systc=
(as defined in IEEE-279), Class II electric syste: (as defined in IEEE-308), engineered safety feature equip =ent, cable pene-trations, or their interconnecting cables required to =1tigate the consequences of the stea= line break accident and place the reactor (s) in a cold shutdown condition; or C
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_ 8-(b) Loss of the ability to cope with accidents due to ruptures of pipes other than a steam line, such as the rupture of pipes causing a steam or water leak too small to cause a reactor accident but large enough to cause electrical failure.
12.
Assurance should be provided that the control room will be habitable and its equipment functional after a steam line or feedwater line break or that the capability for shutdown and cooldewn of the unit (s) will be available in another habitable area.
13.
Enviren= ental qualification shculd be de=enstrated by test for that electrical equipment required to f unction in the steam-air environ-ment resulting frc.. a steam line or feedwater line break.
The in-for=ation required for our review should include the folicwing:
(a)
Identification of all electrical equip =ent necessary to =eet requirements of 11 above.
The ti=e after the accident in which they are require; to operate should be given.
(b)
The test conditions and the results of test data showing that the systems will perfor= their intended function in the environ-ment resulting from the pcstulated accident and ti=e interval of the accident.
Envirensental conditions used for the tests shculd be gelected frem a conservative evalvation of accident conditions.
(c) The results of a study of steam systems identifying locations wher'.
barriers will be required to prevent stea= jet i=p in g=ent frc= dis-abling a protection system.
The design criteria for the barriers should be stated and the capability of the equip =ent to survive within the protected enviren=ent should be described.
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_9 (d)
An evaluation of the capability for safety related ele <;tica; eq uip men t in the control room to function in the environren e that may exist following a pipe b reak accident should be provided.
Environmental conditions used for the evaluation should he selected from conservative calculations of accident conditions.
(e)
An evaluation to assurg that the onsite Dewer dis tribution system and ensite sources (diesels and batteriasl vill remain cperable th rou gho ut the event.
14 Design diagrams and drasings of the s team and feedsater lines including branch lines showing the routing from centainment to the turbine building should be provided.
The drawings should shev elevations and include the location relative tc tne pipin g ttns of safety relatad equipment including ven tilaticn equipment, in takes,
and ducts.
15.
A discussion should be provided of the potential for flooding of safety ralated equipment in the event of f ailure of i feedwater line or any other line carrying high energy fluid.
16.
A descriptien should be provided of the quality centrol and ins;ection p ro z rams that will be required or have been utili:ed for piping systems outside centainment.
17.
If leak detection equipment is to be used in the proposed modificaticns,
a dis cuss
- cn of its ca, abilities should be provided.
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, e 18.
A su==ary should be provided of the amargency proceduras that would be followed after a pipe break accident, including the auto =stic and =anual operations required to place the reactor unit (s) in a cold shutdown condition.
The esti=ated times following the accident for all equip =ent and personnel operational actions should be included in the procedure su==ary.
19.
A description should ha provided of the seis=1c cnd quality clasai-fication of the high energy fluid piping syste=s including the stea=
and facdvater piping that run near structures, systa=s, or ce=penento i=portant to safety.
20.
A description should be provided of the assu=ptions, =ethods, and results of analyses, including stea= generator blevdevn, used to calculate the pressure and te=perature transients in ce=part=ents, pipe tunnels, inter =ediate buildings, and the turbine building folleving a pipe rupture in these areas. The equipment assu=ed to function in the analyses should be identified and the capability of syste=s required to function to =aet a single active ce=ponent failure should be described.
21.
A description should be provided of the =echods or analyses performed to de=anstract that there will be no adverse effects on the pri=ary and/or secondary contain :.nt structures due to a pipe rupture outside these structures.
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