ML20090B436
| ML20090B436 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 03/19/1982 |
| From: | Mattson R Office of Nuclear Reactor Regulation |
| To: | Thomas Young NRC |
| Shared Package | |
| ML082380335 | List:
|
| References | |
| FOIA-91-106 NUDOCS 8204070659 | |
| Download: ML20090B436 (38) | |
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HEMORANDUM FOR: Tolbert Young, Jr., (.hief, Reactor Operations P s gg Section 2 FROM:
Roger J. Mattso'i, Director, Division of Systens Integration
SUBJECT:
CLARIFICATION OF A STEP IN OUR PRELTHIMARY EVALUATION OF OPERATOR ACTION FOR GINNA SG nlBE RUPTURE EVENT
Reference:
Mmorandum dated March 2,1982, from Tolbert Young, &'.,
to R. Mattson The remarks presented in the January 28th memo compared the actual operator actions to the actions recomended in the Westinghouse generic guidelines.
The guidelines recomend that the operator depressurize the reactor conlant system to the ruptured steam generator pressure at soon as the initial cool-ing stage is over (i.e., the core erit tahperature is about 50'F below the ruptured SG saturation temptrature).
Previous background infonnation ac-companying this recomendation Indicates that Westinghouse has been aware of the possibility of upper head voiding. The background infomation for the generic guideline, meant to be used for operator training, states:
"If no RCp is running, voiding in the upper head may occur (depending on the ruptured steam generator pressure) during depressurization of the primary to the ruptured steam generator pressure. This will re-siilt in a rapidly increasing pressurizer Itvel as water displaced from the upper head, in addition to excess safety injection flow, replaces vented steam in the pressurf zer. In addition, depressurization of the primary will be slowed. Consequently, pressurizer level may approach off-scale high before primary pressure decreases to the ruptured steam generator pressure. Pressurizer level must be monitored to prevent filling the pressurizer."
The Ginna event progressed as predicted in the text above, except for the stuck open PORV, which resulted in completely filling the pressurizer and RCS pressure decrease about 100 psi below the ruptured SG pressure.
As mentioned in the mmo, the SG D pressure was lower tbn expected by Westing-house. Thic lower pressure appears to be because the operators started to de-l crease the secondary pressure before the ruptured SG B was isolated. We would expect that if the guidelines had been strictly followed (Reinder: not the actual Ginna procedures), the pressure in SG B would have stabilized to a value above 900 psi (in most other Westinghouse plants, well above 1000 psi).
However, we should recognize that tripptrig the reactor coolant pump may leave the upper head suf ficiently hot thof flashing would take would take place at l
event higher pressures.
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Tolben Young, Jr. gpl9 g The Westingaouse generic guidelines are still under our review and, in a set of questions sent to Westinghouse Owners Group, we have asked that the possibility of flashing in the reactor vessel upper head during the depassurization be further considered. We have not ruled out the possibility that further evaluation by Westinghouse a4 the staff may confim the viability of this portion of the generic guideline. If this would be the case, a note mentioning the possibility of flash < ng may be added.
We appreciate the cassent you have made and agree that the step you have proposed may reduce the potential for lifting the steam safety valves, at the same time limiting the amount of flashing in the reactor vessel upper head. However, the principal goal of depres:vr12ation is to terminate the primary-to-secondary leak to prevent flooding of the steam lines (which is likely to have taken place in Ginna). The importance of this goal has to be taken into account when considering the risks related to a steam bubble in the upper head (which will disappear when the RCP is restarted).
The reactor vessel upper head voiding has been seen also during other events and discussions with tho licensees on the subject are underway, for your infomation we enclose a generic letter sent to all licensees of operating PWR's, if you have any questions about our response to your memo or you would like to discuss this subject further with us, please do not hesitate to call me or members of my staff.
M 5tanet Ruser J. Nat%
Roger J. Mattson. Diredfor i.7 Division of Systems Integration
Enclosure:
As stated Distribution l
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'O -tt LICECIIS Of CPERhilNG FRf 55UR12E0 WALER NUCLEAR POW APFLICANi$ TOR OPCRATING LICCf.",ES (ExCtPT FOR ST. LUCIE, UNIT NO.1)
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SUNfCT:
fATL'RAL CIRCULATION C00tDcwN (Crneric letter No. 81,01 )
On June 11, 1930, the St. to:te Plant, Unit Nc. 1, wa forced to cool down on natural circulation es a r esu'it of a co~poncnt cooling uter r.alfunction.
During the tooldctn process, abnormally re.pid incrf<ases in prasuriier level were observed.
Suhtquent enalysc5 have confirmed that these abnceal level increases are produced by fleshing cf liquid in the uppa bred of the reatter vessel, forcing water out of the vessel and into the pressurinr.
A tore complete denription of the event and circumstances involved is provided in the orclosure which includes a letter sent to the TWR N555 vendors <.oliciting their opinicns and curcrents en the significence of the event and phenezenon-in gener$1.
Based on our review of the event to date, r believe that core cooling was never lost during the St. Lucie, Unit !Io. 1 event.
That specific event does
-not canstitute a direct safety concern.
We have, hcwever, identified two aicas of concern applicable to ill pressuri;cd water reactors requiring prorpt actirn:
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1.
The l',aAcy,tabifi,ty,'CwcuTitToMEjMs 671Tlrsiij,77fa ys of, Vessel Veiding During Anticicated Cooldown n55,1,o n t - lAtural
.r s of RSAlt,c2 Cooldctn with a significant steam void in the vessel requires controlling a "two pessurizer" system, khich is an undis frable challenge to the
- operator, in fact, we are not avere of any training f acilities (simulators) today which would allcw an operator " hands on" experienc'e in practicing such control Moreovt;r, it is cur opinion that any significant vessel void g picduced during controlled cooldown conditions incrtases the' suscept ibility of the plant to niore serious accidents:.
for these reasont i
i reactor vessel voiding during controlled natural circulation cooldowns should be avoided.
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2-As descrited in the enciesure, sessel voiding at St. Lucie Unit No, 1, vis caused ty the eptretor reducing system prtssure such that the ccrresp;nding seturation ter perature dropped to the te.perature of the relethtly 1iegnant fluid in ihe reattor vessel u;;.er ht ed.
Presently, priatry sysum cooldcnn rates are tesed on vit sel structural integrity ccnsidernions and do not esplicitly ccnsider aciding picNction of shnificent sic em soids in the vessel, l'ar, c s er, c oeldcon r cies are t ased on f ivid h m;er at ures rnc t s ured in t he, r ir.:2ry piping, As the St, Lucie t' nit No, 1 e s t nt ha s s b.n, these cotsured t
- (rato rs ten in f6ct be on the crdci p! 100 dt gre es f thrcr,Lcit or n ;re 1; <.er than ihe ;;;+r t r ed fluid
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t(-l4rature, And, thcrtfore, nut indicat ive of the sc*uret icn pi enure of A
all fluid in the pr nry system.
UrA r cendit iens which re quire c oolc'c..n en o tural c h cule t ion and..hrn ripid dep re s s uriz ot icn is not nea t s 6ry t here ray Le a nc htr of ways. to n td d ru ci.cr v eu e l,o icing, f or co pie, a Ic,< (coldemn rate can be t r ecif ied, c;upled with "hoicing' the plant at intfre.ediete ccnditiens to allt < the fluid in the
- pcr ve uel to equilibrate nith the rest of the pr i o ry, sys t e r.
Fri.c s e r, a cidance of.essel soiding by 1cwer prirary syster. ccolda.n r cles can incr(ete the tire required to echieve shutdern cooling entry conditicns a.nd th.;s incr ene the t we eusiliary feednater is dycnded upon to rcxve decay he at (specifically, fer the less-of-offsite pc.,er case).
Thus,' supplies of ( c ne'c r s a t e-grude a nilii iced-water r:ast te considered if cooldcon tines are citrnded.
v 2.
FailuIe of the Operator to Pave Pricr Kncaledae and Trainirm for This f ent
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The ccuse of initial surges in pressurizer level at St, tecie, Unit No. 1, wcs not irmediately recognited or understood by the optrator.
We attribute inis to the f act that icng-term natural circuletion coldonn under the specific circumstances of the e vent was never explicitly analysed t'y the N5SS vendor from the standpoint of trying to recognize a phenom nen ruch as that which occurred et St. L ucie, Unit No.1 In the St. Lucie cvcot, the operator ultirately reccgnized the ceuse of the.
lesel surces and was eble tc maintain control of the plant.
Our corcern, hc.:ever, is the pou ibility of an operator taking incorrect action in an ef fort t o ccrrect for an untoonn event or unrecopized phcnouna.
We believe that proper procedure
- and training 'an provide the nece< sary -
guidante to tbc operators both to avoid reactor vessel voiding as well as recognize it when, and if, it occurs during controlled natural cir culation Ecidown. We are not sure if such proctdures and training are in place at pressurized wat er reactor f acilities.
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1 Ccnsequently, we request that you promptly review your current plant operations in light of the 3t. Lucie, Unat No. I event and the discussions above and i
implement, as necessary, procedures and training which will cnable operators to avoid (if pessible), reco5nize and preperly re act to rcactor vessel voiding l
during natural circulation cooldewn, tle cenclude that the ections describ(d d ove sbould be cD'picted as socn as 4
they reason +bly can be (i.e., within 6 nonths f or epcrating rc actors).
In eddition, to that we ray determine whether your license should be amended to incorpcratt ;Fese actions as requirccints, li:cnsees of ci; rating pressurized water rcactors are requested, pursuant to $t0.E4(f), to furnish, within 6 rcnths of receipt of this letter, an essessr4nt of your facility procedures and training progrtm with respect to the natters described abcve.
Yeur assessmant should include:
l.
a demonstration (e.g. analysis and/or test) tha'. ccntrolicd natural circule-tion cnoldewn frca operating ccnoitions to cold shutdonn conditions, ccnducted in accordance with your procedure, should not result in rcactor vessel voiding; 2.
verifiestien that supplies of condentate-grade auxiliary feedwater are sufficient to support your cooldc..n rethod; and
~3.
a description of ycur training pr ogram and the provisions of your procedures (e.g., limited cooldcwn rate, response to r!.pid change in pressurizer level) t'iar deal with pr evention or mitigation of reactcr vessel voiding.
W Applicants for cperating licensees are requested to implerient the subject procedures and training and provide the requested assess..'ent within 6 months of receipt of.this letter cr 4 rcnths prior to the staff's scheduled issutnce i
- of its operating license Safety Evaluation Report, whichever is later.
4 Please refer to this letter in your response.
I This request-for inforrnation was approved by OMB under' a blanket clearance number R0072 which expires December 31, 1981.
Concents on burden and duplication may be directe'd to the Of fice of Management and Pudget, Reports Managcri,ent,
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Room 3?03, New Executive Office Building, Washington, D.C; 20503.
S ncerely.
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Division o Licensing i
Office of Nuclear Reactor Regulation J
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t.UG 12 USD Letter sent to PWR NSs$ Vendors:
Westinghouse, Combustion Engineering and Babcock and Wilcox Dear lir.
St'BJECT:
V01D FORMTION IN VESSEL HEAD DUR1kG ST. LUClE NATUttAL CIRCULATION C00LDDh'N EVENT CF 6/11/80 On June 11, 1980, the St. Lucie reactor was shutdown due to a lo:s of cemponent cooling water to the reactor coolant pump stbis.
This also" required shutdown of the reactor coolant pumps and c:oldeun was accc~.plished by na tural circulathn.
At approxir.ately 4 hcurs into the event, charging flew, ubich s.as initially being divided between the cold legs and the auxiliary pressuriser spray, t'as diverted entirely to the auxiliary spray to enhance the dcpressuri:atien and reduce the system pressure on the rump stels.
At this time, ttnorr. ally rapid increases in pressurizer level were observed which could not be explained by the charging flew rate alone.
Detailed evaluation and follcw-up analyses by the licensee and N555 supplier have indicated that
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a steam void was probably formed in the upper head region of the reactor ve ft1 and displaced wate. from the vessel into the pressurizer.
Cont'nued alternating realknment of charging ficw between the cold legs and auxiliary spray line uced a " sat:-tooth
- pressurizer level behavior.
Relevant informatien and o.a available to the staff to date are provided in the enclosure.
It has been postulated that the Atcam void in the upper vessel tas produced when the system pressure dropped below the saturation pressure corresponding to the temperature of the fluid in the upper head.
Because the measured hot and cold leg temperatures at the tirne of voiding were highly subcooled *
(~2000F), it appears that the fluid in the upper head was much hotter, relativel stagnant, and in poor cc=unic6 tion with the fluid exiting the core and in the up ?er plenum.
In addition, stored hcat in the upper head structures most liiely contributed to the voiding.
' Because of the unexpected occurrence of the void, !be failure of the operators to inc.edtately recognize the void formation and take corrective action, and the cuestion of whether such void for.T.ation is properly accounted for in safety LH-AOg:(
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h analyses (Chapter 15), we have sent s list aWhting cur concerns to the licensee.
These questions a# the enciesure for your information.
f 1.'e are presently evaluating the need to pursgically with all PUR licensees.
Prior to taking any defipr. we are soliciting your technical opinion and advise ptial for void forr.ation under similar circumstances ing you.
Specifically, we need to know if you can just mg phen */.enon cannot occur in fl5SS's designed by you (or caA pcner.cna can be properly predicted by your transient mend if it can cooldeun occur),is properly accounted for in operatir rates operator guidelines, and operator tr he simulator)
The urgency of this tratter requires,)ou advInden (15) working days af ter receipt cf this letter sMpel inferr.ation submittal by you on the subject 1: auld precivameditiev51y pursue this issue generically with your cusha d
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Paul S. Ogirector for Plant her Division g t'on Office ofmegulatten
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d GINNA STATION B-STEAM GENERATOR NRC MEETING MARCH 23, 1982 AGENDA o
INTRODUCTION i
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INbPECTION AND EXAMINATION RESULTS o
DAMAGE MECHANISM EVALUATION i
o RECOVERY PROGRAM j
o TECHNICAL BASIS FOR REPAIRS i
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PLANT SCHEDULE i
o CONCLUSION 1
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i o-GINNA STATION B-STEAM GENERATOR NRC MEETING MARCH 23, 1982 i
, OBJECTIVES i
o DETERMINE FULL EXTENT OF DEFECTS AND LOOSE PARTS o
DETERMINE FAILURE MECHANISM (S) t o
RESTORE STEAM GENERATOR TO A CONDITION WHICH IS SAFE TO OPERATE MAINTAINING RADIATION EXPOSURES AS LOW AS
, REASONABLY ACHIEVABLE o
OBTAIN NRC CONCURRENCE FOR RETURN TO POWER i
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GINNA STATION B-STEAM GENERATOR NRC MEETING MARCH _23, 1982 PURPOSE OF MEETING o
TO REVIEW RESULTS OF INSPECTIOto TO DATE o
TO OBTAIN CONCURREJCE WITH STEAM GENERATOR PROGRAM CONCEPTS o
TO OBTAIN APPROVAL FOR REMOVAL OF STEAM GENERATOR TUBE SECTIONS e-t t
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e GINNA. STATION B-STEAM GENERATOR NRC MEETING MARCH 23, 1982
!{S_ARB/NRC REVIEWS o
concurrence with program concepts NSARB - 2/26 NRC
- 3/1 o
dpproval of removal of metallurgical samplec NSARE - 2/26 NRC
- 3/1 o
approval of repair program NSARB - 3/16 Nhc
- 3/23
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o approval of return to power NSARB - mid April NRC
- late April 4
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GINNA STATION B-STEAM GENERATOR NRC MEETING MARCil 23, 1982 INSPECTION UPDATE NO. 4 WEDGE AREA 1
o R45C54 g"
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R44C56 p
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r G 11f!4 A S T A T 1 011 B-STEAM GEliERATOR 14RC MEETI!1G MARCli 23, 1992 METALLURGICAL EXAMI!1ATIOlj o
site photography o
Westinghouse R&D laboratories o
model for wear orientation comparisons o
photography at 90' increments o
radiography at 45' incrementn
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o' transverse cross sections of column 55 tubes
-R42C55 - 2.5" and 4" from upper end
-R43C55 - 2.5" and 4" from upper end
-R44C55 -
2.5",
4" and 8" from upper end
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1 GINNA STATION B-STEAM GE!1ERATOR NRC MEETING 1982 MARCH 23,.
e LABORATORY EXAMI!!ATION REMOVED TUBING e WEAR SUFFACES e PRIMARILY RUBBING WEAR CIRCUMFERENTIAL DIRECTION e NO EVIDENCE OF CORBOSION INVOLVEMENT e EVIDENCE OF SURFACE COLD WORKING e FATIGUE STRI ATION ON FRACTURE SURFACE e TENSILE OVERLOAD BURST TUBE FAILURE SURFACE
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MECHANISM EVALVA1: 0N DROGRAM e
INVESTIGATION OF VARIOUS INFLVENCES MECHANICAL LATERAL LOADS GROSS FLUID 8.OADS AKIAL LOADS tGCAL FLUJD LOADS 8
HISTORICAL INFORMATION REVIEW 6
INITIAL PERIMETER TUBE INVESTIGATION 0
LABORATORY EXAMINATION OF REMOVfD TUBE SECTIONS O
E00EL TESTING 0
LABORATORY COLLAPSE AND FATIGUE TESTING 9
FIELD TEbTING AND EXAMINATION l
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GINNA STATION B-STEAM GENERATOR NRC MEETING MARCil 23, 1982 I
LABORATORY EXAMINATION REMOVFD TUDING 9 WEAR SURFACES e PRIMARILY HUDDING WEAR CIRCUMFERENTIAL DIRECTION 9 NO EVIDENCE OF CO"ROSION INVOLVEMENT 9 EVIDENCE OP SURPACE COLD WORKING e FATIGUE STRIATION ON FRACTURE SURPACE 9 TENSILE OVERLOAD BURST TUBE PAILURE SURFACE 4
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I GINNA STATION B-STEAM GENERATOR i
NRC MEETING i
__ MARCH _23, 1982 CORRECTIVE ACTIONS i
l t
o eddy current examination t
o video inspections i
o-obtain metallurgica) samples
-o remove structurally cagraded tube sections o
restore preventatively plugged tubes.to service
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o eddy current examine tubes adjacent to repairu i
secondary side video inspection following repairs o
1) primary and secondary hydrostatic tests metal impact monitoring system l
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i GINNA STATION l
B-STEAM GENERATOR NRC MEETING f
MA.R.CH 23, 1982 I
REPAIR OPTIONS l
e o
EDM cutting process a
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mechanical cutters o
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o remove from tubesheet end I
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GINNA STATION 1
B-STEAM GENERATOR
-CATEGORIZATION OP DEFECTS-1 2
NO. 6 R OC70 NO. 4 NO. 2 CATEGORY WEDGE AREA'
< REA WEDGE AREA WEDGE AREA I
l.
Structurally R8C92 R42C55M R43C59 R44C55M Degraded R11C91 R43C53 R43C60 R44C56 R12C91 i
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March'23, 1982
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HYORAULICS CONSIDERATIONS:
I
- TUBE FATIGUE DUE TO FLU 10 INTERACTIONS i
- FLUID-E!ASTIC STABILITY
- TURF 0LEP1C 4
- LOCAL FLUID EFFECTS - EDDYS, CRACK 1
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STRUCTURAL CONSIDERATIONS:
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- FIXED-FIXED BOUNDARICS
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VIBRATION AMPLITUDE DUE TO VORTEX SHEDDING AND CROSS-FLOW TURBULENCE ARE RELATIVEL( UNAFFECTED BY SMALL O!STORTIONS AND SURFACE IRREGULARITY I
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I EFFECT OF HOT LEO TUBE REMOVAL ON CROSS-FLOW VELOCITY
( %-D MATHEMATICAL STUDY )
i INPUT CROSS-FLOW VELOCITES:
6.0 ft/sec. hot leg 5.4 ft/sec. cold leg AVERAGE VELOCITY IN MAX. GAP VELOCITY IN CASE TUBE REMOVAL REGION FIRST TUBE R0W DOWNSTREAM NOMINAL 5.30 19.40 10 TUBES REMOVED 3.69 13.50 40 TUBES REMOVED 3.86 14.14
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TUBE REMOYAL DOES NOT ADVERSELY AFFECT FLOW VELOCITIES I
e
FATIGUE EV ALUATION OF SURFACE DMuGED PLUGGED ' TUBE (ASSUME FULL AXIAL RESTRAINT AT FIRST TSP) 9 ENVELOPING TRANSIENT - PLANT LOADING / UNLOADING. 14,500 CYCLES 8
ASSOCIATED LOADS e
TEMPERATURE VARIATIONS PRIM ARY TH0T:
547F (HOT STANOBY) TO 602F (100% POWER)
SECONDARY TST:
547T (HOT STAN0BY) TO 518F (100% POWER) e EXTERNAL PRESSURE RANGE:
795-1020 PSI e
AXIAL TUBE LOAD RANGES e
TUBE-10-SHELL THERMAL MISMATCH: + 780 lbs.
ASSUMPTION - TUBE IN CONTINUOUS THERMAL EQUILIBRIUM VITH SECONDARY FLUID; STUB-BARREL WITH INFINITE THERMAL INERTIA e
PLUGGED-T0-ACTIVE TUBE THERMAL MISMATCH: 0 TO + 1200 LBS.
ASSUMPTION - SINGLE PLUGGED TUBE WITHIN A CLUSTER OF ACTIVE TUBES 9
AXIAL BENDING LOADS e
AS-BUILT MISALIGNMENT, 0.25 INCH e
TS-TO-TSP THERMAL GROWTH MISMATCH, 0.05 INCH e
TS ROTATION DUE TO PRIM-TO-SEC ap, 0.08 INCH i
FATIGUE USAGE CALCULATIONS _
MAXIMUM STRESS INTENSITY RANGE S
= 55.25 KSI RT 3
(ADJUSTED TO E.= 26 x 10 KSI FOR ASME FATIGUE CURVE)
ASSUMED STRESS CONCENTRATION FACTOR FOR SURFACE DAMAGE = $,0 (MX. PER ASME)
NUMBER OF CYCLES n = 59,000 (THIS NUMBER RE-PRESENTS ALL TRANSIENTS LUMPED CONSERVATIVELY)
MINIMUM ACTUAL USABLEECYCl.ES PER ASME CODE N = 135,000 CALCULATED USAGE = n/N = 0.4303 TuaE Geonerpy
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e-COLLAPSE.INTEGRITYEVA!LJA,Tj0NS BASED ON EXTENSIVE LABORATORY TESTING COLLAPSE PRESSURE FOR NOMINAL TUBING - 5000 PSI COLLAPSE STRENGTH IS RELATIVELY UNAFFECTED BY SHORT (1 i TUBE DIAMETER), THROUGH-WALL TIGHT CRACKS FOR TUBE COLLAPSE, DUE TO THE MAXIMUM ap - 1020 PSI I
REQUIRED WALL DEGRADATION IS - 80%, IF UNIFORM, AND
> 90% IF LOCAL TUBE COLLAPSE RESULTS FROM PLASTIC INSTABILITY AND REPRESENTS y llELOADING CONDITIONS AN INSTANTANE')US FAILURE MODE. 0F FORRGETUBING.THEMAXIMUMop{r1020 PSI)OCCURSDURINGNORMAL r-OPERATION WHICH, THEREFORE, REP T NTS'A PROOF TEST,
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