NRC Generic Letter 1979-45

From kanterella
Revision as of 11:49, 4 March 2018 by StriderTol (talk | contribs) (Created page by program invented by StriderTol)
(diff) ← Older revision | Latest revision (diff) | Newer revision → (diff)
Jump to navigation Jump to search
NRC Generic Letter 1979-045: Transmittal of Reports Regarding Foreign Reactor Operation Experiences
ML031320181
Person / Time
Site: Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Nine Mile Point, Palisades, Indian Point, Kewaunee, Saint Lucie, Point Beach, Oyster Creek, Cooper, Pilgrim, Arkansas Nuclear, Prairie Island, Brunswick, Surry, North Anna, Turkey Point, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Duane Arnold, Farley, Robinson, San Onofre, Cook, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Fort Calhoun, FitzPatrick, 05000273, Trojan, Crane  Entergy icon.png
Issue date: 09/25/1979
From: Ross D F
Office of Nuclear Reactor Regulation
To:
References
NUDOCS 7911260004, GL-79-045
Download: ML031320181 (71)
v • d • e


  • ' ti- rth EGastw UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON D. C 20555Sb -3As-SEP 2 5 1979TO ALL POWER REACTOR LICENSEES

SUBJECT: TRANSMITTAL OF REPORTS REGARDING FOREIGN REACTOR OPERATINGEXPERIENCESThe enclosed reports are provided to you for information and use inyour reactor evaluations in light of the Three Mile Island Unit 2accident. Enclosure 1 is an internal Westinghouse report which describesan incident involving a stuck-open power-operated relief valve thatoccurred at the Beznau Unit 1 reactor in Switzerland on Augsut 20, 1974.This report is now a part of the official records of the President'sSpecial Commission investigating the TMI-2 accident. Enclosure 2 is aninternal NRC staff memo on this incident. Enclosure 3 is a report on asteam generator tube "rupture" incident at the Doel 2 nuclear power plantin Belgium.If you have any questions about the enclosed information, please letus know.D. k'Ross, Jr., DirectorBulletins and Orders Task Force

Enclosures:

1. Technical Report on BeznauUnit 1 Incident of August 20, 1974:TG-l Trip/Reactor Trip/SafetyInjection Actuation2. Memorandum dated May 15, 1979; AshokThadani to D. F. Ross, Jr.3. Memorandum dated September 13, 1979;Darrell G. Eisenhut to Multiple

Addressees

.1,4/q 1)lg(lOovcf640

.i -4 jTo : O.A. Wilson (with att.) : T. Cecchi(3 copies) Date : tSeptember 4, 197.cc : F. Noon (with att.) Ref : SA/251H. Cordle (with att.)D. ten Wolde (with att.)A. Hall (with att.)T. Currie (with att.)'J.P. Lafaille (with att.)R. Galletly (with att.)R. Lehr (with att.) Pitts.J.D. Mcadoo (with att.) Pitts."A. Weaving (w/o att.)W.B. Thee (w/o att.)* R.L. Cloud (with att.) W. Rockenhauser (with att.)SUBJECT : TECHNICAL REPORT ON NOK 1 INCIDENT OF AUGUST 20, 1974References (1) Telex SE-G-74-195 (8/26/74) to NOK by-H. Cordle(2) Letter (8/27/74) NKA-3940 from L. Barshaw.You will find attached the technical report on NOX I Incidentof August 20, 1974 prepared by WNE inspection team who wentto Beznau on August 23.This report, which should be sent to Beznau, summarizes ourobservations on the course of the transient, the damage aswe viewed it, our calculations and conclusions.Despite what is indicated in the referenced (2) letter, inorder to have a more complete report, we added some recommend-at.ons for future changes. / T-T. CrrC-CT!-< SYSTEMS ANALYVS; i .: _*i. .-_._ ;:. ., , bc_- .4 .,^.;

a%- /\ Gu SA/251TECHNICAL REPORT ON BEZNAU UNIT ONEINCIDENT OF AUGUST 20, 1974 TG-1 TRIP/REACTOR TRIP/SAFETY INJECTION ACTUATION.J.P. LAFAILLR. GALLETLYT. CECCHI ,H. CORDLE, Director,Svstems FnnineerinaSeptember 2, 1974 DISTRIBUTIONH. CORDLEA. HALLD. ten WOLDE0. WILSONL.--BARSHAWT. CURRIER. GALLETLYF. NOONJ. LAFAILLET. CECCHIR. LEHRJ. MCADOOR. CLOUDW. ROCKENHAUSER TABLE OF CONTENTSTECHNICAL REPORT ON BEZNAU UNIT ONE INCIDENT OFAUGUST 20, i974 : TG-1 TRIP/REACTOR TRIP/SAFETYINJECTION ACTUATIONPaceI. INTRODUCTION., -1II. SEQUENCE OF EVENTS 1III. TRANSIENT BEHAVIOR OF MAIN PLANT VARIABLES 3IV. DAMAGE TO THE PIPE RESTRAINTS AND SUPPORTS 5.V. EVALUATION OF THE INCIDENT 7VI. OTHER RECOMMENDATIONS 14VII. APPENDIX A 16VIII. FIGURES (18) 20.I

.I--l 1T -INTRODUCTIONThis report is produced as a result of a site visit followingthe incident on Beznau I which took place on August 20, 1974.The object of the visit was to make a rapid evaluation ofwhether the consequences of the incident would jeopardize safety.This report confirms the telex of Aua. 28, 74 on this subject.The scope of this report, therefore,- is limited to a descriptionof the sequence of events and of the damage observed togetherwith a Dossible explanation and assessment of safety issues.It is not meant to be a corprehensive analysis of the effectsof the incident.-...* _.II -SEOUENCE OF EVENTS DURING THE INCIDENTl .IOn Aucust 20, 1974, a trip of one of the two turbines on theBeznau I reactor followed by failure of the steam dump system,.to operate resulted in a reactor trip and the opening of thepressurizer relief valves. One of these valves subsequently.failed to close and the extended blowdown of the pressurizerresulted in the rupture of the pressurizer relief tank,burstinadisk.' Exarnination following the incident revealed that thepressurizer relief valve which had failed to close had beendamaged, as had some of the supports to the pressurizer relief 'line itself.The sequence of events, with times where known, is reconstructedbelow :Initial conditions :Date : Aucust 20, 1974 Time : 11.20 a.m.Pressurizer pressure : 154 bar Pressurizer level : 50%Pressurizer relief tank level : 80%Power outnut of turbooenerator 1 : 187 tVW (e)2 : 177 MWV (e)

-2 -TimeEventDisturbance occurs on the external gridnetwork.TG1 trips out on high casing vibration.11 *hrs 20 min 07.8 sac Vibration causes low A p signal fromhydrogen seal oil system.-+ Steam dump valves fail to open./ SG steam pressures rise.Pressurizer pressure rises.Pressurizer level rises.20 11.9 Both pressurizer relief valves open.20 17.3r- -Turbotrol of TG2 drops into the emergencymode.20 23.0 One pressurizer relief valve closes inaccordance with automatic signal,pressure continues to fall and levelcontinues to rise.Pressurizer relief tank pressure rises.Pressurizer relief tank level rises.TG2 power level falls then rises to anoverpower of 214 MW (e).21 00.4 Reactor trips on pressurizer low pressure.21 01.2 TG2 trips.SG steam pressures rise. ISG water levels fall.Pressurizer level falls.23 03.5 Secondary side safety valves lift.23 13.9 Steam is formed in the ACS hot legs andpressurizer level rises past 100% andremains off-scale for 3 to 5 minutes.A reasonable assumption is that waterdischarge occurs through the open reliefvalve.Operator shuts pressurizer relief lineisolation valve. (Reported verbally as2 to 3 minutes after the trip).... ..I.1. /

  • S--3-Pressurizer level falls rapidly as steambubbles in RCS collapse.Pressurizer relief tank bursting diskruptures.Pressurizer relief tank pressure falls.Pressurizer relief tank level falls.11 hrs 23 min 43.5 sec High containment pressure recorded(peak 1.1 bar abs.).24 51.2 High containment temperature recorded(53.4 C).25 17.8 High. containment activity recorded(17.3 mr/hr).32 14^.3-*-- *-SIS initiated as pressurizer level fallsto 5%. LPressurizer level rises as SI water isadded to the RCS.SIS stopped manually.Subsecuently Procedure begun to bring reactor tocold shutdown condition using the atmos-phe:4o steam reliaf valves.Fig.. 18 shows the record of pressurizer pressure an d leveltransients following incident initiation.; , ' ., * *. .*- .I B S b .l *III -TPANSIENT BEHAVIOR OF MAIN PLANT VARIABLES DURING TEE INCIDESA turbine trip in a two turbine plant is equivalent to a 50% loadrejection and no reactor trip should be initiated if controlsystems work correctly. Since in Beznau I the steam dump systemdid not work at all, initially the main variables behaved asfollows :1. Steam Generator steam pressure rose (to about 66 bars) butnot enough in order to actuate safety valves.2. Feedwater flow, stebm flow and steam generator level decreasednormally as expected.

3. The reactor being in automatic control, the nuclear powerdecreased. When reactor was tripped after about 49 seconds,it was at 76%.4. Pressurizer pressure rose rapidly from 154 bars to a maximumof 160 bars (pressurizer relief valves actuation) in about11 seconds.5. Reactor coolant system average tcmpcraturc rose rapidly froin298.5iC to a maximum of 305.5*C in about 50 seconds,6. -Cold leg temperature rose rapidly from 275°C to 2906C, thendecreased to 240'C in 10 minutes, to 2200C.in next 100 minuteand to 140'C in next 170 minutes.,7. Pressurizer level rose from 50% to 67% in about 50' seconds.Due to the fast pressurizer pressure increase, both pressurizerrelief valves were rapidly actuated. Their actuation took placealmost simultaneously. However, it is very probable that thevalve actuated by the compensated pressure error signal (signalelaborated by a PID controller) opened some seconds before theother one due to the derivative term of the PID controller.When pressure decreased below relief valves actuation setpointthe valve directly controlled from an uncompensated pressure'signal did not shut. This resulted in a depressurization at ratEof about 0.75 ba-r/sec, resulting in a reactor trip by low pressu:in approximately 49 seconds.The reactor trip signal tripped the turbine which was still inoperation, resulting in a further steam pressure increase (above70 bars) which produced steam generator safety valves actuation,lowerinc -he pressure to about 65 bars.../. ..

-IReactor coolant system average temperature decreased to about285'C and pressurizer level to 23% in about 1 minute afterreactor trip. At this point pressurizer pressure had fallento hot leg saturation (70 bars). Subsequently, hot leg flashingresulted in an increase of pressurizer level until the pressurizerfilled about 3 minutes after reactor trip, resulting in probableliquid water'discharge from the relief valve and bulk boilingin the core. x Then the operator isolated the failed relief valve,and pressurizer level decreased reaching the setpoint (5%) forsafety injection actuation (safety injection is actuated bycoincident low nressurizer pressure and level S.I. signals) about11 minutes after reactor trip. The system then started refilling.When pressurizer-level reached about 70%, safety injection pumpswere shut off manually.The reactor was then brought normally to cold shutdown conditions.IV -DAAGE TO TH.E RELIEF PIPE RESTRAINTS AND SUPPORTSFor pipe layout, see isometric, fig. 1 attached.The relief line to the power relief valves comes out of thepressurizer top and runs directly down (vertical run of 6.8 m).'It passes through a grating floor. No impact evidence betweenthe floor and the pipe insulation exists. (Gap about 25 mM).At the bottom of the vertical run there is a console typerestraint. (Location 1 in fig. 1). The main dimensions aregiven in fig. 2. There is contact evidence, as shown on theficure, but no damage.The pipe then runs horizontally to the restraint 2 (fic. 1)This restraint limits motion of the Poie in a hor'zontal direction,neroendicujar to the pipe axis (See fig .3). Scratc;hes on theshoes ind.cCate that the pipe moved about 26 FM axiay. m The toppart of the insulation is slightly swishes (See fig. 3).* * / *

  • I3: Nuclear Pcwer was to.--. P. -..

I -.,The line then runs vertically down (2.77 m) and separates intotwo branches each having a stop valve and a relief valve.Fig. 7, 8 and 9 show the damage to the valve.Examination of the pressurizer relief valve which failed toclose revealed that the yoke had broken off completely. Onearm of the cast iron yoke had broken at the top and the otherarm at the bottom taking part of the voke ring with it. Thetop break showed the presence of a very large flaw (inclusion).All broken faces showed classic brittle failure together withevidence that the faces had rubbed together following failure.In addition it was reported that the valve spindle had beenslightly bent. This was not observed since repairs had already.been started.Fig. 6 and 7 show the pedestal of the support between.the twovalves. Fig. 4 is a sketch of the support and details thedamage.The damage corresponds to a rotation of the pipe around ahorizontal axis perpendicular to the pipe axis. No evidence oftranslation has been found. Considering fig. 7, the back bol _swere strained much more than the front ones.The bolts of the undamaged valve support have been inspected.It was found that -the paint was cracked at the bolt joints, butno other damage could be found.After the valves the two branches of the pipe drop to thelower floor. Fig. 10 shows the penetration corresponding tothe damaged branch.At the lower floor, the restraint R4 (See fig. 1) has been pulledoff the floor (see detail in fin. 14). The motion has beenimposed on the frame by the bar of the hanger passing througha 50 mm slot in the frame (See fic. 11).../ ...

'Sn --Pestraint R5, which is onlv a column supporting a sliding shoe,shows a motion of 70 mm as shown in fiq. 5.The pipe then j.oins a header and passes through the floor (R6 or.fig. 1). There is evidence of 25 mm upward displacement.At the lower floor the header has an elbow. Motion is restrainedby a snubber. The bolts fixing the snubber to the concrete'*'ere found to be loose.V -EV.AwLUATION OF THE I!CIDENITThiS evaluation covers the incident transient effects and apreliminarv estimate of magnitude and probable causes of damageto the pressurizer relief vinina and supports.1. Comvarison with desian transientsThis Beznau I incident is similar to the two following incidentwhich are normally considered among readtor coolant systemdesign transients :-Loss of load (up to pressurizer relief valves actuation).-RCS depressurization (from Pressurizer relief valvesactuation).* *From the standpoints of core power, heat transfers and systemspressures and temperatures, the reported incident is less sever'than the desicn transients considered above.The magnitude and variation rate of the temperature and pressuretransients resulting from the incident are indeed fully coveredbv the values used for equipment design.Plant variable behavior durina the transient did not result inan uncontrolled or damaging si:uation, and the released activity

-8 -T~ g'remained well below dangerous lim~its. All existing protectionsystems (steam generator safety valves, reactor triD, safetyinjection) worked properly and were adequate to handle theincident avoiding core and equipment damage.2. Evaluation of damace to the Pressurizer relief line, therelief valves and suonoorts.The relief line between the pressurizer and the power reliefvalves is part of the reactor coolant pressure boundary andtherefore is important to.the safety of the plant.The one poster relief .valve which failed to close was isolatedin accord with design intent by the operatcr closing theappropriate relief isolation valve and hence no uncontrolledloss of coolant occurred.The review of the relief line equipment showed damage to therelief line supports and the pressurizer relief valvePCV-456.The damage evaluation and probable causes are treated below.a) _Discussion of the incident related to cause of damaae.Examination of the relief line and supports along with therecords of primary reactor coolant system parameters leadsto the following observations.(1) It is probable that the observed damage to the sunportsis the result of hydraulic shocks from a sequence ofwater and steam discharge through the relief line.(a) The pressurizer relief line from t;.e relief valveto the pressurizer can fill with condensate. :hisdistance is apprcxiratelv 19 meters, and can conwaina voluzne of 0.06 m'. Openinc of the relief valves../. .

will cause a rapid discharge of the water. Theresulting dynamics are one Possible cause of thepiping displacements observed.(b) Based upon the recorder chart of pressurizer waterlevel, it appears probable that some water discharg;occurred later in the transient when the pressurize:was completely filled. .The records indicate thatthis event could only have occurred after automaticclosure of the undamaged valve (PCV-455C).Dynamics related to this event are another possiblecause of the observed piping displacements andsupport damage.(2) It is not possible from available evidence to provideone sequence of events which uniquely explains theobserved results of the transient.It is not certain that the valve damage was theconsequence of the same hydraulic shock that resultedin the support.damage.The observed sequence of events indicates that onelikely scenario is as follows :(a) The undamaged relief valve, PCV-455C, opens firston the derivative compensated pressure controllera few seconds before the second valve opens.(b) The water slug formed by condensed pressurizersteam in the relief line is largely dischargedthrough the undamaged valve. We note that thisportion of the line sahowed little or no su=mortdamage.

I , Ir(c) The second valve, PCV-456, opens on continuedpressure increase and the transient, combinedwith the large flaw in the valve yoke results invalve failure.With this hypothesis, there is no reason to expecta hydraulic shock higher than in opening of thefirst valve hence pipizg displacement sufficientto damage supports miaht not yet have occurred.(d) The first valve closes automatically upon a reducincpressure signal before pressurizer water levelreaches 100%.(e) Water discharge occurs upon filling the pressurizercreating a substantial hydraulic shock in the reliefline. Since the undamaged valve has already closed,the resultant pipe displacement was most pronouncedin the portion of line where the damaged valve islocated.Other scenarios can also be postulated, but none hassufficient support of evidence to permit identificationof a single sequence of events as the cause of observeddamage.(3) The events which lead to corpleze filling of thepressurizer and the second water discharge throuchthe relief line required more than a single failure :(a) The failure of all the secondary steam dump valvesto overate.(b) The failure of the pressurizer relief valve toclose. It is likely that such a failure would not

-11 -/have occurred even with an initial hydraulic shockwithout existence of a larqe flaw in the reliefvalve yoke.(4) Considerina the valve PCV-456 itself, when in the openposition, there is a spring force producing a tensionof about 60,000 to 80,000 xewtons in the yoke. W-henthe disk lifts, this force can be anplified due todynamic effects. The presence of the flaw in one ofthe arms overstressed that arm (area reduction andstress concentration), which caused it to break.This caused a moment to be applied to the other arm,resultirn in beri4ira of the spindle and rupture.of the base.. The broken retal surface anpearance wastypical of brittle failure with some polishing due to.rubbing contacts following o7okP se arat~in. The yoke t.-rose about 2,5 cm, the normal stroke of the valve.with the broken voke, the valve failed to close.Dynamic forces due to the free motion of the operatorbody may have contributed to damage to the support...(5) Appendix A calculates the forces and stresses on the-relief line piping in two locations, suspected to beamong the most stressed. It is seen there that, withinthe calculation assumption the piping could have beenmarainally overstressed. However, since a dye penet-anmcheck of the PVC-456 valve to pipe weld was reportedto show no defect, we cannot see any reason to thinkthat the plant would operate in unsafe condition with.the line in the present sta.te. This statemen- assumesof course that all the support sxystem Of t-.e pipingwill have been returned to its design condition beforethe reactor goes back to pcwer.

-12 -To gain further assurance on the safety of theline we would recommend that a dye penetrant checkof all welds near the fixed points be made atthe earliest convenience. The locations includethe pressurizer nozzle, the relief tank nozzleand the intermediate supported or restrained points.b) Ooerational Considerations(1) Plant operation with one pressurizer power relief valveclosed off does not present a safety problem. The highpressure reactor trip and the pressurizer safety valvesprovide the necessary protection against overpressureof the reactor coolant pressure boundary.The-existence of the power relief valves is to preventunnecessary opening of the main code safety valvesduring certain plant design transients.(2) The safety injection system functioned normally with,a reported total injection rate of 40 1/sec. Theinjected water raised the pressurizer level from 5% to75%. Assuming the injection water to be initially at16'C and atmospheric pressure in the RWIST and to end upin the pressurizer at 285°C and 110 bars then thequantity of water leaving the RTHIST must have been about10 M3.This would cause a decrease in ?WST level ofabout 0.7%. The injection time would be about 4.1/2minutes assuring a constant injection rate.../. ..

'-13 -(3) The reason why the turbotrol gear of turbine 2 droppedinto the emergency mode is not known. It was reportedthat the effect of this would be to lock the turbineinlet control valves in their last position. Thus the-would no longer respond to changes in steam pressure.This pay account for the overpower excursion experienceon turbogenerator 2 just prior to its tripping.! (4)!The failure of the steam duzrp valves to open wasreported to be the result of a wrong wiring connectionwh-ch was not -iscovered during testing. The controlcircuitry of the steam dump valves had been out formaintenance at some previous date. Before being putback on line, the circuitry had been tested in twohalves. Each half was checked independentlv'of theother half an6 each half checked out satisfactorily.A fault at the interface of the two halves thusremained ur.revealed...- .*i! I I* * ;'e,* I .

1' oVI -oTHER RECOlMMENDATION:S1. The piping displacements and support damage which occurredhave indicated the possibilitv that the Pressurizer reliefline was marginally overstressed. The likelihood is thatthe displacements resulted from either discharge of a waterslug initially in the line or from relief of water when thepressurizer was corpletely filled.The initial evaluation of stress was deduced from observedsupport displacement and support bolt strains. As such, nodefinitive indication of possible stress levels with thistransient exists as basis for ad~ evaluation of fatigue damagefor the entire piping length.We would recommend a dynamic analysis be performed, consideri:at a minimum the effects of the steam condensate initiallyin the line. The force time history function can then be usedfor evaluation of fatigue damage as well as the adequacy ofrestraints.2. The failure of the power relief valve yoke is more probabledue to the use of cS~t-- onmateriads of Q~sruction whereimpact Properties are poor and flaws of the type involved inthis failure can remain undiscovered.We therefore recommend such non-destructive tests as arefeasible be made to ascertain that no flaws of this type existin the valve currently installed.Further consideration might be given to replacing these yokeswith a less brittle material.../. ..

/-3. The test procedures followino maintenance of the controlsystem to the steam dump valves should be rewritten toeliminate the possibility of unrevealed faults.4. It would be useful to provide means (i.e. 2 separate alarmsone actuated bv the uncompensated pressure signal and theother bv the compensated nressure error sional) in order toknow if certainly each pressurizer relief valve opens durinaa pressure excursion., ,...*5 ' .-.SI. I,;% .If , I .

II -APPENDOX AStress and Force Evaluation ih the nine between valves1. Darace to the sunoortThe two bolts on the right sidce on figure 3 were strainedabout 3 mm. The two bolts on the left side were alsostrained but only to the point of getting loose.2. Evaluation of the moment aonlied to the sunnortBolt size : M10 -Shaft size' (diameter)8.888 < d < 9.128 mm(Cataloaue MARC-GERARD -1970)Section (average) w (8.888 + 9.128)2 63.73 mm2;T 2,Assume for the bolt material a yield stress ofa 32 ka/mm2Hence the moment to strain the two bolts isM -63.73x32x2x.135 -550.6 kg.m3. orce -ecu4red to create that moment.~.f l ~38;5t TlVA L *EP ?C1 456 5LAL 33/1 460 405 _ _ 00i I j-17 -A-2If one neclects the effect of. the supports located downstreamof valve 456, one can write the ecuation385.F = 135%R.1Knowino that R x.135 = 550.6 kgmHence F = 1430 kaIt is felt that such a force is in the possible ranne.4. Stresses in the nip2 (Primary stresses only)Pipe : 3" sch 160Hence : OD = 3.5 in = 88.9 mmt = 11.13 mmBending modulus = v= 47.17 10 mm3Bendina stress :-~ 32a = M 550.6 10 = 11.67 kg/mm2B Pe/r s s 47.17 1tPressure stress (ASI'E III, Article NB 36 52)ap = vOD -164.5x10 2 .88 .92x 11.136.57 kq/mmCombination (Article NB 36 52)P1D + 2 2I iB1 and B2 are taken from table 3683.2-181 = B2 =Hence1= 6.57 + 11.67 = 18.24 kaimn20tot

-18 -A-3'5. Allow'abl._tQeSA 376 Grads_ 316S at room tuma. = 20 ksi -14 kalmm2Sm at 6501' (-3430C) -16.6 ksi = 11.6 kg/mmAllowable stress = 1.5 S (ASME III, article NB 36 52)* m21.5 S = 21 ka/Irm (room temperature)M2= 17.4 kg/mm (343°C)6. Conclusion for orimarv stresses in the pineSince it annears that hot fluid has been carried by the pipefor a time of about 3 min, the hot allowable stress needsto be taken. Then it anpears that the actual stress isslichtly hicher than the allowable18.24 > 17.4 ka/rrm2It should be noted that the fieiure of 18.24 k/zm 2 is aminimum, since it corresponds to the plastification of thesupport (M = 550.6 kar).7. Primarv and Secondarv stresses in the mineThe evaluation of secondary stresses (article NB 3653.1)recuires the knowledge of the temperature gradients inthe pine. It was thus not possible to evaluate thesestresses.8. Primarv stresses at the reducerBending xontentBe = 1430s (385 -(405 -13N) .; rm.= 357 kcrm 1 t: -A-4reducer 21 " sch 1CC£OD = 2.875 in = 73.02 mm t = .375 in = 9.52 mm.I3 3I = 1.64-in = 26.9 cmPressure stress = rOD = 6.28 kd,/mm2Bending stress = = 13.28 ka/rn22Total stress = 19.56 kg/mmThis stress should be considered more as indicative sinceit depends 6o much on the assumption of the force location.The same conclusion holds as for the pipe stress.

  • 1-,. ,,Z'I -_" I.J..Vi_ I.1 .II,.I"$4he El.S c- .: f- Ie tv.;,; W ritrLf, II 1,4L ;/I /1A6LL7(. 2).N.%A-I!l- I,*II..;_IDirection of arobabl'.-\efort.ABolts (6 total): Hexagonal head = 25 nunDamace : -no general distortion-no rubbing evidence-contact evidence in AFigure 2 -Restraint R-1

, AX>V4 VA w*1II'LOWView A/n(mrk o thc a 6 shos)-D KeeZ r, crt(marks on the shoes)Damage : -top of insujation slightly smashed-scratches on shoes as shown on view AFigure 3 -Restraint R-2 iIP.C.IiI1, '!.4 .III Ik, -_. .-Bolts 1 (4 total) m-102 (4 total) ti-13 ,4 total) pull out.-rce = 41'/hclt*Damaae: -no evidenceat straps, pipe andbolt- (I) ar.d (3)-all 4 bolts (2) .ave been strained-gap measured as shown-strain evidence in the r profile as 3YFiaure 4 -Restraint R-3II

-ri- P.. C4 LQ N X'.. Ut ,,1 .11. Q )Figure 5 -Restraint R-5Motion Evidence.33 I ' on1-dEZNAU -UNIT No 1 (NOK) v-'STEAM DUMP FAILURE It4CIDL:;4TAug. 21, 74PRESSURIZER RELIES LINE .Figure 6 -Undamaged Relief Valve.

a IsB- : * : :: I ..0 : ,)STEAM DUMP F;_URE INC 'DE:TAug. 21, 74PRESSURIZER RELIEF LINE.5"; bj '- 4. -1.;-.5-- --7ej1o*'1FtFigure 7 Damaged relief valveGeneral view showing the two fractured armsand the liefted operator.11'1i3- ISTEAM DUMP FAILURE LNCIDrE-JTAug. 21, 74PRESSURIZER~ RELIEF LINE* S ..' ,Z. ~& .:"~ '-. -.:.A.4 0Figur Daae VleI De.ail of.fqaIsuu.,yoke BEZIJAU -UNIT N* 1 .(NOK)STEAM DUMP FAI1.URE INC I )ENTAug. 21, 74PRESSURIZER RELIEF LINE; v b .A d-}~~~~~~~~~ .*4*' A, .-'";_. '-:,';9 i.¢ ---s I. w A9. S. ._ _ :. s. ..-*.. ;- v>l :I~ A.*iur 9 Damaged' Valve-eal .fatue rbne.2.

\,LZNJAU -UNIT :-, I * .~~\STEAM4 DUMP FAILU1.,,. INCIDENTAug. 21,j 74PPLESSURIZERJ RMTJLIe LZNE.Of../jj.?~e* ,. , ~ *t*A./N! I A. .**1Figur 10.-Elbo afe daae vaveI a

.j jB1EZNAU -UNIT No I (ISTEAM DUMI' FAI J.' iIE INC TDL:'U'Aug. 21, 74PR^ZSSUR1ZE1I 1ULLIEF LINEFigure 11- Support R4 (1)General arrangenmcnt100 x 50 x 5 profiles50 Dun s lott. on vi .

' :IMU -UNIT WZ I (Nw ;STIMAM DUla' IAILU1W INC 1DUlNT', , JAug. 21, .74PRE.-.:URIZLR RUELICEF LINE.~*E -,i .-; 'tw,. _e/;A mJ-. -2

  • r ***,AAC : ..P , 'C t.1~
  • An ' P~- e.1 ICFigure 12 -Support R4 (2)I"Y BEZNAU -UNIT N0 1 (?.STEAM DUMP FAILURE INCI Do'tNTAug. 21, 74PRCSSURIZtR RELLEF LIIE.. rg*?-. ..*aZ4.a. I..* ...., _Attachmen to loLItE" '*,,* t [ii. Ib,: r -!_I.^ ._ i * -.* r /tt J ' -F ,  ; ' ,* ' .. ...;". -.; .
  • 1.Iiur *1 -.Su**j*port I4(1ta1 m n to 'l oConcrete damage (back cf the restraint.)

! %NAU -UN'l' No 1 (NQK)-STEAM DUMP FAlLURE INCIDENTAug. 21, 74PRESSURIZER RELIEF LINE.W. -. I.'. V.N.\S"IFigure 14 -Support R4 (4)Detail of concrete damage.A....

II IBEZNAU -UNIT No 1 (NON)STEAM DUMP FAILURE.INCIDENTAug. 21, 74PRESSURIZER RELIEF LINE.IFigure 15 -Ceiling Penetration (1)(q-3 TBEZNqAU -UNIT N* 1 (NOX)STL1AMl DUMP FAILURE I:NCeIDENTAug. 21, .74PRESSURIZER RELIEF LINE.Figure 17- Ceiling Penetration (3)Kci)

Irt X X 1 ;.- -- .-- -.-3i-- '--\ i-\.\- !--'-i~-*1{ ___-.- -4 ... .-. ;II -'-. ! ..' !i, *1 m ---~~; '-- 1 I5-'-1 --*ii--4 -*-1 4 .4_. _ .,.... __. ,_.' t i- -1 _ _i1 °'.,, ; _ _; , .... _..- -- s-~.- -1 .... ---. _ I.. .\ , ...:4_f ollo.ig incident i f.4.tiagion.C(R) 9&bk I INOX RFPORT ON P-uZNAU AccDr)ENtT 0F AUGTJST 20, 1974TRIP TG-1/REACTOR TRIP/SI/On Aucust 20, 1974 at 11:20 a.m. a trin on turbine TG-1occurred resultinq to high bearinq and casinq vibrations(Bearing 6:60 )At trio time, generator 2 was delivering about 140 MVar.Resulting from a failure of the steam dumn system tooperate, with the consequence that the relief valve didnot open. That resulted in a rapid rise of coolanttemperature, steam pressure and pressurizer level andpressure.At 160 bar of pressure in the primary, the Pressurizerpressure relief valves opened, lowering raPidlv the Pressurein the orimarv. About 10 seconds after valve onening,the oressure had reached such a low level that the pressur-izer pressure relief valves were reactuated to close. Dueto a disturbance, valve PCV-456, failed to close, resultinqin a lowering of RCS oressure up to 100 bar after about1 minute. Reactor trinned resulting from a low pressuresianal (126.5 bar).Due to the openina of the pressurizer relief valve, thepressure in RCS drooped to about 70 bar, corresponding to'a saturation temperature of 284'C. Consecuentlv, steamappeared in the primary hot leg, filling the pressurizer.Two or 3 minutes after trip, the operator recognised thefailure of the relief valve and isolated it with the poweroperated valve 531. The water level began to dron, and11 minutes after trip, automatic SI was initiated by lowpressure ann level in the pressurizer.

1 4 .I !'Pace 2. I.SI systems worked normally and about 40 litres per second ofwater was soilled through the four SI pumo nozzles into theprimary, causing a rise of pressure to 110 bars and afurther rise of level to 70 %. The SI Pumas were thenturned off and the Dower operated valves of the soray pipinqswere closed.From that moment on, the pressurizer level could be controllethrough charging pumps and release of steam, assumini the.orimarv to cool down.About 3 minutes after trio, the containment oressure alarmsignal was actuated because of too high Pressure, and 1minute later the high activity alarm. Maxim=m pressure incontainment reached 100 mbar over normal. The operatorsactivated the containment fan coolers. Since severalsafety alarms of the pressurizer relief t~ak were on, it wasquickly assumed that the rupture disc was brokIen and thatthe discharge channel was defectuous. After TG-1 trio,due to steam dumn failure, steam pressure rose to 66 bar.The turbatrol of TG-2 was actuated as an emergency afterTG-l trio. TG-2 was unreaular in behaviour, and thePosition of the control valve retained constant during thepressure transient. The oerformances of TG-2 rose to about214 MWe due to higher steam pressu-e (rise from 52 barto 66 bar).After TG-2 trio, following reactor trio, steam Pressure roseto over 70 bar, actuatina the safetv valves and thus lowerinaPressure to about 65 bar..2. C(TROrOhLGICAL OST~N(~ F OV SAugust 20, 1974 Paae 3.2.1. Reactor TrioBeginning of incidentTG-l main breaker offPressurizer nressure low-tripReactor trip breaker openTG-2 main breaker offSI actuation (pressurizerPressure and level low)2.2. Events as Recistered on Ai.arm11 h 20' 12"39,7n later39,8' later40,3" later11'55,9" laterTvpoewriter-TIME11:1511:2C11:2C11:2C11:2]11:2]11: 2]11:2111:2111:2111:2111:2:11:2;TG-1 power highAllowable oil pressureof TG-1 too lowPressurizer pressurehigh.Pressurizer pressurehigh.Reactor Trip.Tavq RCS-A hiqh1 Steam nr. upstream ofTG-1 stop valve hiqh.L Tava RCS-A hich1 SG-A steam oressurehich.L SG-R steam pressurehiqh.L Steam or. upstream ofTG-l stoP valve.1 SG-A steam pressurehicih.1 SG-A steam pressurehiah.a Safety oil nressure ofTG-2 too low.2 Tavg RCS-A135,5 MWar158.2 bar159.9 bar302.2*C66.3 bar305.20C67.3 bar67.2 bar77.6 bar73.3 bar65.4 bar11: 2;285.2 eC Paqe 4.T:IMP11:2311:2311:2411:2411:2411:2411:24Steam pressure uostream ofTG-2 stop valve.Pressurizer relief tanktemperature hiQh.Pressurizer levelPressurizer levelContainment oressure hichPressurizer relief tank levellow.Pressurizer relief tank pressurehiqh.Pressurizer relief tank oressureSG-A+3 steam oressures normal.Containment activity highLoop B RCS flow low.Containment air temoerature hiahPressurizer level low.Pressurizer level normal.Surqe line temoerature too low.Pressurizer levelthich.68.1 bar62.86C79 '88 %1.1 bar abs.20.2 %0.59 bar11:2511: 2;11:2511:2611:2711:3211:3211:3311:340.15 bar63.7 bar17.3 mr/h88 I53.4 °C6.8 '18 %271. 1C58 %.2.3. Seauence of :Events for Pressurizer and Pressurizer Relief Tar.TIME11 h 20'11.1"11.9"22.8"23 .0"23 .0"23.1"24.2"33.o"35 .nPressurizerPressurizerPressurizerPressurizerPressurizerPressurizerPres-urizerPressurizerPressurizeroressure above control ranae.relief valve.relief tank oressure hiahrelief valve lcokedpressure normalrelies .ank le-ve' hiahlevel hich.relie-f tank oressure too hich.Dressure under nornal.4..(ZO

.IPaqe 5.TIME11 h 21'00 .4"01.2"05.1"13. 5"11 h 233'11 h 24'11 h 25'27.6"43 .3"43.5"47.1"29.4"51.217.8"PressurizerPressurizerunlocked.PressurizerPressurizerunlocked.PressurizerPressurizerContainmentPressurizerPressurizerContainmentContainmentoressure low -Trio.pressure low -SISrelief tank level hiqh.pressure low -SISlevel hich -1 channel t::relief tank level too him.pressure too hiqh.relief tank level low.relief tank nfessure normatemperature hich.activity hich.3.' A%'qALYSIS OF O'FF CAUSESOF THE INCIDFNTTG-l trioped due to hich casing vibrations, especially incasing 6. It had already been noticed that TG-l wassensitive to shocks. At the moment of incident, TG-l wasset to function under maximum effort, so that it couldsupport a maximum of vibrations.The trio is not unfamiliar and would not have affected theprimary if steam dumr had normally been actuated.An inspection of containment after primary h.ad cooled down,showed that the yoke between the PCV-456 valve housina andair engine was broken, and probablv due to a dynamic efforton the pining at opening of the valve.Consequently, the valve failed to close ar.d imitiated araDid fall of pressure in nrimary. The pressurizer relieftank rupture disc broke, due to a mrolonced surce of orintarvcoolant in the tank. Items 2 and 3 show the disc brokewhen the relief valve had already closed.

I IPaqe 6WATER COLLErC TED IN CONTAINMFE'r SUMPRegen. hold up water Tank A 38 % -100 9.8 m3Regen. hold up water Tank B 16 % -36 -= 3.2 mTotal quanlity of water collected =13.0 m3Pressurizer relief tank 80 % -19 = =11.2 m3Water out of system. -1.8 MSince no further damage was noticed in containment, itcould be assumed these 1.8 m3 of water were blown out.4.1. Thermal Stresses in RCSBeside a rapid water temperature rise of about 6C afterTG-1 tripped, a rapid primary pressure rise fron 154 barto 160 bar, there was also an imoortant temperaturetransient in area of SI nozzles. However, since thereactor's main pumps operated all the time, thus mixinq-cold spray water with hot coolant, it can be assumed thatother components didn't underao high ternmerature gradients.Furthermore, nozzle temperature and stress remained withindesign limits.4.2. Damaaes to Relief SvstemsDuring insp'action in containment after cooling of Drimarv,the following damaces in the pressurizer relief. systemswere observed-relief valve PLV 456 Mechanism broken on both sidesand bent snindle.-One anchor point of the relie' svstem ninin" after valve-Relies cank pressure disc broken. was loose.Further damages in ccntainment were not noticed.

Pace 7It must be said that the relief tank is not desicned toaccent steam from the Pressurizer for a Drolonqed time.The damaqes to the relief valve is therefore a directcause to the breaking of the rupture disc.4.3. TurbinesTG-1The cause of vibrations to the casinc are most nrobablythe stresses and shocks. The P sicnal from hydrogenseal oil svstem is due to casina vibrations.Damaqes to the seal or casing are most improbable.TG-2 aThe oscillation from 172 MLWe to 110 MWe, and then to 215 M-Wesuggested that the bolts of the high pressure cylinder wereloosened and had lost some of their tension.A too small stress was noticed, due to leakaqe of theseals of the high pressure cylinder. Due to too hiqhrotational momentum at 215 MAe, the couplinq between turbineand generator was closely controlled.5. When reviewing the sequence of events, the Failure of twosystems, namely the steam duimb and the Pressurizer reliefsystem, we came to the conclusion that it did not brincto an uncontrolable nor a damaqinq situation. nurina theincident, no activity (in gas or liquid form) in thesurrounding area reached an uncontrollable level.The generator safety valves maintained the steam pressurewithin allowable limits. The SIS broucht back the Primarvto a safer pressure, allowinc normal cooldown conditions.6.' PROPOSAL FOR MODIFICATTONS6.1 Control of cenerator 1Generator 1 reaching ranidclv to casinq vibrations,it will Paqe 8be tried to see if the regulator can be modified in orderto have a quick action.6.2. Pressure ReaulatorTests will be made to see if the first row of impellers inthe pressure regulator of the turbine must not be reviewedin order to limit power to 190 MWe.6.3. Steam DumD Systema) Revisions and calibrations should be made in Ateam duffsystem (before opening of steam dumo valve.)b) Studies will be made, to make periodic controls ofsteam dumo while in operation. It should helo to.insu--.better safety limits (for example : unwanted oneninq c.1steam dump valve).c) A control type writer linked to the steam dumo willbe installed in order to control the opening of steamdumo valves and to check the qood working of oil OUmos .6.4. Pressurizer Relief SvstemThe first measure to be taken, is to reoair the damacedvalve, the pivinc supports and review holticrms.The pressurizer relief tank rupture disc must be remlaced.With these repairs start-um should be possible.To see how the relief svsterm nipina can be better securedand how shock at opening of relief valve can be avoidedare further measures to he taken.

W -UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHWINGTON, D. C. 205-YIS1979MEMORANDUM FOR: D. F. Ross,.Jr., Deputy Director, DPMFROM: Ashok Thadani, Task Manager

SUBJECT: STUCK OPEN POWER OPERATED RELIEF VALVE AT FOREIGN PWRIn the process of gathering data on power operated relief valves (PORVs)for our report on Westinghouse plants, we were informed by Westinghousethat they were aware of only one instance of a PORY failing to recloseafter opening. No failure of this nature had been observed on any U.S.reactor plant designed by W. The failure, according to W, occurred atone of the NOK reactors in Switzerland. Our survey of aTl operating U.S.W reactors also indicates that the failure of a PORY to reclose has notSeen observed on any U.S. Westinghouse reactor.To follow up on the apparent foreign reactor PORV failure, we contactedHoward Faulkner of NRC International Programs and informed him of ourneed for additional information. Our basic need was to determine whetherthis failure did indeed occur and, if so, if It could occur on a U.S. PWR(due to similar system and component design).A phone conversation between NRC (H. Faulkner,.Ashok Thadani and ScottNewberry) and the Swiss Federal Office of Energy of Switzerland was ar-ranged for the morning of May 15 to obtain this information. HowardFaulkner informed the Swiss that we would treat this information as con-fidential and would telecopy them a copy of what we intended to includein our W evaluation report prior to its issuance.A sequence of events for the turbine trip and associated PORY failureto close described by Mr. F. Weehuizen, Head of Energy Section, is attached.We requested additional information to supplement that in the phoneconversation:1. Event reports pertaining to the event2. PORV description, manufacturer and failure mode71 i/2=oo73 IU. F. Ross, Jr. 2 -2 -1979Based upon this phone conversation, we note that:1. As demonstrated by this event, pressurizer level will remain abovethe trip set point for ECCS actuation for a stuck open PORV.ECCS did not actuate automatically until the operator shut the PORYisolation valve.In this case we do not know how soon the coincident signal (Lo Level/Lo Press.) would have automatically initiated HPI and the subsequentoperator actions since the PORY was isolated atiminutes.2. The indications in the control room of actual PORY position andrelief tank parameters appear to have provided the operator withsufficient information to make a reasonably rapid assessment of theproblem and take appropriate action.Since this event occurred about five years ago and because of itsrelevance on our current deliberations on W designed plants, werecommend that complete information package including plant databe obtained and reviewed, as well as the role of the operator.We therefore recommend that all operating Westinghouse reactors modifythe pressurizer level/pressure coincidence ECCS actuation as alreadydirected by I&E bulletins 79-06 and 79-06A and that we continue to pursuethe PORY design, manufacturer and transient sequence to make a determina-tion as to the likelihood of this event on a U.S. PWR and to obtain moreinformation on turbine bypass system failure modes as a lower priorityconsideration.A. ThadaniTask Managercc: E.G. CaseR. MattsonL&L. TedescoXT. Novaklf7.FaulknerS. Newberry Enclosure1. Trip of 1 turbine due to generator disturbance(plant has a twin turbine arrangement -only 1 turbine tripped -no direct reactor trip unless both turbines trip)2. Secondary system pressure increased -turbine bypass(5 relief valves to condenser) did not open due to acontroller malfunction caused by operator error duringprevious maintenance period.3. Primary system temperature, pressure and pressurizer levelincrease. PORV opens.4. Primary pressure decreases. After 10 seconds PORV should have shutbut remained open.5. Reactor trip on low pressure(pressurizer level still above low level trip, thereforeECCS has not yet actuated on coincident low pressure -low level)6. Reactor Coolant System pressure decreases to saturation. Voidingin hot legs. Operator observes flow oscillations and reactorcoolant pump vibrations. He did not trip the reactor coolantpumps.7. 2-3 minutes after the reactor trip, the PORY isolation valve was shutby the operator. He had received increasing pressure and temperatureindication in pressure relief tank. He also had open indication ofPORV (direct from limit switch on valve stem) in the control room.8. High containment pressure alarm ('a1.4 psig).High containment activity (pressure relief tank rupture disc ruptured).9. Pressurizer level decreased. 11 minutes after the reactor trip, ECCSactuated on coincident low pressure/low level ECCS performed as designed10. Pressure increased to 110 bars ('1600 psi).Pressurizer level increased to 70% of indicated range.Operator tripped HPI and maintained pressurizer level usingcharging pump (CVCS).11. No core uncovery.No fuel damage.No hydrogen generation.Additional Notes:1.. Main feedwater was maintained throughout the event.2. Secondary system reactor trips are:-low steam generator level-both turbines trip.©g 3. Total reactor coolant last to contaimnent suinp

  • 1.8 cub~c meters...a.

mitt fli,<inMEMORANDUM FORiFROM:

SUBJECT: UNITED STATESNUCLEAR REGULATORY COMMISSIONWASHINGTON, D. C. 20555SEP 13 1979H. R. Denton, Director, NRRE. G. Case, Deputy Director, NRRD. Ross, Deputy Director, DPMR. Mlattson, Director, DSSDarrell G. Eisenhut, Acting DirectorDivision of Operating ReactorsINCIDENT AT BELGIUM DOEL 2 REACTORIn response to our following up on a rather large, sudden steam generatortube rupture at the Doel 2 nuclear power plant in Belgium, we havereceived the attached report. You may find this incident particularlyinteresting since the unit underwent a transient where pressurizer levelapparently went offscale high. Strip chart recordings of the event areenclosed.We hope to be obtaining more informa ion on this event in the near futureDarrell G. Eisenhut, 4cting DirectorDivision of Operating 'Reactors

Enclosures:

As Statedcc: S. HanauerF. SchroederB. GrimesP. CheckG. LainasS. LevineV. StelloW. Russell0

_.Q CENTS0 DETUDE DE L'ENERG/ NUCLEAIRE-,C.E.N. I S.C.K.MR Jim 0;ar .. tW *ug..'Yd .uW6euSrVeuiI'ei adressr votre fipones Mr. Joseph D. LAFLEUR, Jr.on doux examplsres aua Deputy DirectorLABORATO0IES DU C.E.N /S.C.K. Office of International ProgramsBouretang 200 B. 2400 MOL UNITED STATES NUCLEAR REGULATORY COMMISSIONle. 01e X 1e (0p 2 WASHINGTON D.C. 20555elex SCKCEN-Moi 31922 SEP 19793L U S ACAdr. twegr.: Centratom Mol I. U.S.AMOL. is 21.08.79.V Ilnre V/rnfI. Nrif.Centrale BR3FM./mb5.5126/71

Dear Dr. LAFLEUR,

As a first answer to the telex of Mr. H.J. FAULKNERNRC-BHDA, dated 8.8.79, I send you here enclosed a report describingthe steam generator leak incident at the Unit 2 of the Doel nuclearpower plant.This report has been transmitted to me by "TractionelEngineering", a division of the compagny "Societl de Traction etd'Electricite" in Brussels ; as you most probably know, this divisionis playing the role of engineering office for the benefit of the Doelplant operator compagny (EBES).I hope you will find in this report satisfactoryanswers to all your questions ; do not hesistate to ask for eventualadditional informations.Yours sincere .F. MOTTEBR3 Plant Superintendent.Enclosure : "Report on the incident at Doel 2 nuclear power plantSevere leakage in steam generator B on June 25, 1979".t10 PD/vEF20.07.79SEP 3 1979REPORT ON THE INCIDENT AT DOEL 2 NUCLEAR POWER PLANTSEVERE LEAKAGE IN STEAM GENERATOR B ON JUNE 25, 1979.1. STATUS OF THE POWER PLANT AT THE MOMENT OF THE INCIDENTThe primary system was being heated up after repair worksat the actuation system of the main steam valve.At the moment of the incident, temperature in the primarysystem was -2551C (refer to point A on Fig. 1 & 2) andpressure had reached its rated value of 157 kg/cm2 (referto point A on Fig. 3 & 4).Tne reactor was subcritical with all rods in.Secondary pressure in the steam generators was -45 kg/cm2the saturation pressure corresponding to 255'C (refer topoint A on Fig. 6 & 7).For some time, A-loop steam generator had shown a lowactivity value along the secondary side (below admissiblelimits) that indicated a small leakage.2. SEQUENCE OF THE EVENTS (refer also to various computerdata given in attachment)2.1. Initiating phaseAbout 7:20 PM, a quick pressure decrease is recorded in2the primary system (about 2 kg/cm per minute : see Fig. 4),which.results in accelerating the operating charging pump.A second charging pump is started manually. The letdown©CD 2.station of the CV system closes automatically. It is confirmedthat the relief valves are closed and their isolation valvesare preventively closed. The level in the pressurizer quicklydecreases (see Fig. 5) and the electrical heaters areautomatically disconnected.At the same time, a quick level increase is recorded.inB-loop steam generator (see Fig. 7 point B). The activitymeasurement channels of the blowdown system record a maximumvalue.The combination of all those signals indicates a severe leakagein B-loop steam generator. The faulted steam generator isthen immediately completely isolated along the steam sideand the discharge valve to the atmosphere is set at maximumpressure.Meanwhile -the third charging pump is started (was set apartto be maintained) , but the three charging pumps are notsufficient -to compensate the loss of fluid in the steamgenerator. Indeed, the CV tank is readily empty and thecharging pumps are automatically supplied from the 2R11refuelling water storage tank. To increase the subcoolingprimary pump B is stopped and letdown starts througb A-loopsteam generator (see Fig. 3, point B).2.2. Actuation of safety injectionAbout 20' after the incident started, the threshold pressure2(118.5 kg/cm ) to actuate the safety injection is reached.The emergency diesels start within the required time lapsebut are not necessary. Phase A isolation and ventilationisolation of the reactor building are achieved. The vitalcomponents not yet in operation are started.

j .When reaching the 108 kg/cm value, all HP SI-pumps dischargeinto the primary system, and the pressure decrease isstopped (see Fig. 3, point C).To prevent the secondary pressure in the faulted steamgenerator from reaching the opening pressure of the safetyvalves, the primary pressure is successfully decreased (seeFig. 3, point D) through maximum spray in the pressurizer(re-start of primary pump B and use of both spray lines).During this phase, the level in the pressurizer quicklyincreases and it fills up completely (see Fig. 5). Spray-is temporary stopped and pressure stabilizes at zero flowpressure of HP SI-pumps.The automatically started auxiliary feedwater supply results ina pressure decrease in B-loop steam generator (see Fig. 7,point C). The auxiliary feedwater supply pump of the faulteddisconnected steam generator is locally stopped and isolated(Fig. 7, point D). This cannot be performed from the controlroom since the SI-signal stillprevails. The auxiliary feedwatersupply tank is filled up from Doel 1.2.3. Cancelling of SI-signalPressure decrease was now mandatorya) to avoid the opening of safety valves of the faultedsteam generator.b) to start, as soon as possible, the shutdown coolingsystem (low pressure circuit 1) to stop the letdownof slightly contaminated steam through the A-loop steamgenerator.

4.Firstthe safety injection signal had to be cancelled.This had to be performed more than once (each time requiring5 minutes interval) because of a relay fault.After definitively cancelling the SI-signal, two HP S%-pumpsare stopped and soon thereafter a third one (Fig. 3, point F).While considering the subcooling margin, the last HP SI-pumpis stopped. Pressure successively decreases to reach -65 kg/cm2 (Fig. 3, point H) (saturation pressure is 4 15 kg/cm2at that moment).It is then tried to initiate the CV-discharge line, butvalves do not open. Some time goes by before t-hereason therefore is determined. Due to phase A isolationthere is no longer a-compressed-air supply in the reactorbuilding. After re-opening the compressed-air supply linethe discharge'line is opened (Fig.-3, point I). Pressuredecreases, first quickly, then slower._ .....'.The loss of compressed-air supply has also resulted in theclosure of CC-valves to the primary pumps. The pumps haverun for a long-time without cooling of the thermal shield,however without alarm temperatures were reached.2.4. Initiation of the residual heat removal systemAs the CV-system permittted only a slow pressure decrease,X 15- S.the interlock, which maintains the isolation of the RHRSup to a pressure of 28 kg/cm , has been bypassed at 31 kg/cm2There was indeed a sufficient margin compared to the design2pressure of the system (42 kg/cm ). Thanks to thisoperation the letdown through A-loop steam generator couldbe stopped earlier and the discharge of slightly contaminatedsteam could be reduced (Fig. 3, Point J).2.5. Further sequencesThe abovementioned operation allowed a primary pressuredecrease below the value of secondary pressure in the faultedB-loop steam generator. The secondary level decreases, whichcreates a dilution risk. The boric acid concentration iscontrolled every half hour (stabilized howerver at + 1500 ppm).Thanks to the cooling down, pressure decreases slowly inB-loop steam generator and reaches a value lower than theprimary pressure. From this moment on, attention is paidto always maintain the primary pressure higher than that inthe steam generator.Despite the cold water so discharged in the steam generator,pressure goes on decreasing slowly (due to the presence ofa warm water film at the water surface).As the level of water in the steam generator approaches theupper limit of the broad level measurement pressure issufficiently low (+ 12 kg/cm2) to inject nitrogen.The secondary drain line is coupled with system B for liquidwaste, and the steam generator discharges into It throughthe nitrogen pressure.The nitrogen is only slightly contaminated after thisand can be discharged via the annulus between primary andsecondary containments.

I 1iA, '=6..2.6. Comments and conclusionThe incident has been handled as proscribed and no damageshave occured to the environment or the installation.The procedures have to be reviewed considering the following a-cancelling of phase A isolation to restore compressed airsupply in the reactor building.

Attachment 1 -Computer data1. Initiating phase19 21'06":9 22'51"19 23'31"19 23'32"19 25'42"19 26'14"19 30'30"10 30130"19 38'32"2. Safetypressurizer pressure below reference pressuredemand for charging pump higher speeddisconnecting pressurizer heaters by low levelCV letdown station valves closedclosing of isolation valves of relief valves andspray valveslow pressure in primary systemvery low pressure in pressurizerhigh level in B steam generatorB primary pump disconnectedinjection phase19 40'18"19 40 '19"19 40'19"19 40'19"19 40'20"19 40'24"19 40'33"19 43'28"19 44'39"19 53'1219 56'37"19 57'11"19 57'29"19 58'48"low pressure in pressurizersafety injection through low pressure in pressurizerdiesels startedreactor building ventilation isolationphase A reactor building isolationactuation signal HP SI.-pumpsHP SI-valves openedvery large auxiliary feedwater flow to A SGvery large auxiliary feedwater flow to B SGauxiliary feedwater supply pump B disconnectedvery low level in auxiliary feedwater supply tankpressurizer level normalpressurizer heaters re-startedhigh level in pressurizer3. SI-signal cancelling phase20 00'15" automatic starting signal of diesels cancelled andSI-pumps starting signal cancelled20 00'21" back to SI 2.20 03'24" LP compressed air in reactor building20 05'59" safety injection ordered20 06'05" safety injection20 10'59" reactoF building ventilation isolation ordered20 21'15" HP SI-pump B disconnected20 25'22" HP SI-pump A disconnected20 38'33" valve CC 096 closed20 40'25" valve CC 099 closed20 48'54" compressed air supply to reactor building restored20 49'00" primary pumps CC-valves re-opened4. Actuation of RHRS22 35.54" valve RC 003 opened180

  • i'; -l j,- ~ u .'S X B -_:;-A j 'WAA%FIGUUR 1Isch) i ver:,1. 131. RC 052, Ro. RC 25.2 A -I RC 2Lou AvaQ ,elus A warm beenlus B vrarm been; -350cC0 -3500C.-I TI .lI 9R O I I 2-,-I.been0.50.0 ' iI ..I-_ .2Ro. '2 R 25 1j I.1 A ;~ zi~. .i1,,2zt, (10, l .I :2 t.'..7vio.riiiiei : -30-350°i S.....i .- I _JIIc I°C .*. 3 .II; i :.l :-__ I :_ I _.2I0 250 ..3 fl3J-* -__ _ __ __ __ _ I I-. -: ..,' -2Ii .,I.I , .! ,II.1I--,I I ---;.7* I II I: i -Si. ' 2 RC05LuSA 1-1 be: ;' jD0- IRo p A~:AC5_u kI *'1 n Mt ._-t l.cnln2~C...i..*.............iI ~ t0 2 20 .300 350__________ __________*Il l Bi. y W- _u:)i j I_2 R.O. 2 FC 25ILUSA, vo.a pvulW -ZJV r__ i IJ~ I;J.1 'S ' -LuS B wo fn bcen-I .j!0-3c~ooCj0I1Z,-_----AI A1 -c-ocII.IIII I -l. j eTEE2ZRATU KOUD BEENt.FIGtULR 2i* Schrijver I1. I1. RC 092. Bo. RC 292. A -1 RC 11lu A koud beeA'lus' B koud beeno -350°Co -350°C2'.-.. -.A- -.. -- ---.III.I11III.4III(f0)

% C3go~RU ECFIGLTR 3Schrijver : 2 A -1 RsC 31. Bl. iRC 11 druk RC,2 eg0 -200 kg/czr2*S.,...aI*I.I~.I]II.'Ii.i.IIIiI0..alI I IQ ID RIA A 2 FIGUV; 4ISchrijver : A A -1 PA 2LIIIIa1. 3B. PR6 -7 -8 -9 Druk R2dienst spoor :115 -175 kcg/cmn2Meting in__ _ _ll7,-,- -'I1.W,II ..1 1 .; j'.I I'I. .1!.;Ili,.It'. .I.. .,@!, .A.1 , I 'k !I !-I IIillI075Al!Ipt .!I I I.1I.1 6....II ,I TNii7, 1.L.I! I1I I ..111:;..11li l!II! II1II.IIIII..1 * !I II I175II;hm1~I Ii, ..'I-S-I1.l.II14-! I .I HII 1.I4IIIII.* .1I I! ' ' III ..LI:I'Iil.i !:I,III..-OAp1*II I I__- ---.-.---flTT________ ______ 11.111!IIIIIi.--;D.A! en- j7?Rj,-7j 8, V 'Drvk 2 R ?, :U5-.175!kg/Fm,2; PEIL R 2 FIGUUR 5Scnrijvcr : .2 A -1 PR I1. Bl. L. PR 71 -12 -13 Peil R 2.2. Ro. L. Ref. Ref. Peil R 2.-. .O -100 rO -100 %E'eting in -dienst spoor 1:.@,III.._i.., II .IITII'II:.,1.1 I .I;! l ;.Ii I j,!iiIiIIIIIIITlT IT,1"I,,; Ij rI, ~I"I I:II 'h ' ~ i iIQ L_. _l I : -I.T I-- I I

  • 1***.._.I',',l;'i~lildj',' ,/ 1t4II:! h !. 'I! i!I' :1 i:i'`1l I CI v2j{ti r ~ ;~V0tfA: t..; ____ il~--0T111 .! S4 ~7i1 11!!!ll1n*,~ I^Ji !i; *~ ~ Ji'^lu.,'xi:HI! hII.IhIIISthI!'I jk.* verkeerde sc haalvervangen docor.,--I O+ 100 %^ULosqx 11 Cc-,.d %C--ct1o -lc>o 7C.I) .IC73)_, IjliI, :I.1I I -I. I: i a -.SO. A PEIL -DRUK FIGUMT 6d.0Schrijver : Z A -4 FW Y. 'S4 4 &I. 31. L. TW 9A -10 A Peil SGA.Cq 4, A2. Ro. P. S 4 A -6 A DruX SGAO -3500 mm0 -85 kg/cm,2.Meting in dienst spoor 1 :spoor 2: 0iII .t.IIII ' ,:.I 1. 1i :.,:,.Ll. ...,II.. Il.a II..i ! !I ;... ....._I .;i r .. 4A :1'1I 1.u i_.; ! ki.i%IISI._ft SliiI* 'I **Irv..It I I*l i IAr' I
  • I S g I
  • I...2 *I .11I I
  • I.I
  • S.I I,ft Ift I I* ' ft I* '.5 5 i I i i ISISI I ftI 6Sri IS lft*IS Ijig. I5;' jIllIiI II aftp.-I I:* I1 ftI I i ft I'I'IIIIiII,I ivI, !,; i*1 PI ij ;i- Ii * ...* ._ , ......_fiILI;...1I , *II5"I" kI,,.....ISIi : I IIii. ; !i I 'iI I IIIIIIIiIIJI1a:I ..SI!:,I .., ji ;.iII.! a.tI.....S7Dx ISC~ 8 -&XL~A,-~ lan--SG. B PEIL -DRUKFIGUUR 7Schrijver : 2. A -4 FW YM 51. Bi. AL. FnW 9 B -10 B Peil SG Bcc., .2Pjt 6 D4d i c2. Ro. P. HES 4 B -6 B DruX Sd Z0 -3500 am0 -85 kg/cr2meting in dienst spoor I :.spoor 2 :IIIIIIII.1.IIIIIIIIIIIII.IIIIIIIIIIIIII/1'iI1.-1.Ii A'RC ;?hl hr~ -o' 1 7, 'WAEO71 (CO40)(1-096C92442)fPO 08/30/79 0640ICS IPmIHA I l.3SSIISS Ff, 'Ul 30 O640P?;*S O.iY COV'.iISSIOi\ '. ASHiNGTON1 OCU '! 9.1! 3 FI 3 ?2 1 TG3 /229lU'J.'!W- CO BEAN 13?8.NTW^RP~ TELFX 132/125 30 1017 PI/S.R.iR JOSEPH D LAFLEURJ JR)FPLJTY D)IRECTOROFr1CF Or I\TFRNATlON:AL PROC-RAMSU STATFS NUCLEAR .F:G-JLATORY CO-.X-ISSION-WAiS;~INCTONIO). (-20555)COOPL¢;j:ENTMAAY TC t',Y LF.TTER REF 5951P6/71 OF AUnUST 21 PLEASET; F ¢FT-i SPE.CIFIC Ai'St'!E..S TO ThE FOUR OJE3TlONS kRAISL RYYOU.R ). F4ILcNF:R ON THrlF .JOFL 2 STEANM GENFic~ATOR I.NCIDEtN'T1s TiRE MAC-N1TULDEOL (Q0555) 5.516/71 21 2 1.3/229 -:R JOa'?. 0 LAiL-. .J.7 JN Pe/59oT i--- LFAk ::.As ESTIMAT-D AS .49OUT 30 TONS/eOU.. ANr-)V: L'j? RAOPIflLYS. TF-E LEak IS LOC.RTEL, ON THE TOP OF PIPE NR 1/?hs OF STEA^. GENER.ATOR3 007L ° IN Thr -TRA--OS OF Th' U-$ENli)3. 3UoP.CTED CA'JZ' STRFSS-CORROS10N DU;:. TO OV4LIZATIONh. !rENTINP MAXIMyUM 450 MICR9NETFA.RNCt rLOWCOL 3) S. 1/?4 3o. A. 450.31/?9 il. JOSFPI-. D L47L7UR JR P3/25SLOT DEFOR ATION AT ALL NO TU?EF i:ALL THMNINJ FOUNflTHESE ANSVERS, WERE FO.Ri;ULATEFi .3Y ThE DOEL I .AiND °PLANT SUPERINTENDENTYOURS SINCEP*ELYF MOTTECOL 1 q.RETPl Mr'St:N Nv,.U Tl'.X '.!SAHC. --'C 0Mr. William J. Cahill, Jr. 50_3Consolidated Edison Company of New York,_Jnc. 50-247cc: White Plains Public Library100 Martine AvenueWhite Plains, New York 10601Joseph 0. Block, EsquireExecutive Vice PresidentAdministrativeConsolidated Edison Companyof New York, Inc.4 Irving PlaceNew York, New York 10003Richard RemshawNuclear Licensing EngineerConsolidated Edison Companyof New York, Inc.4 Irving PlaceNew York, New York 10003Anthony Z. RoismanNatural Resources Defense Council917 15th Street, N.W.Washington, D. C. 20005Dr. Lawrence R. QuarlesApartment 51Kendal at LongwoodKennett Square, Pennsylvania 19348Theodore A. RebelowskiU. S. Nuclear Regulatory CommissionP. 0. Box 38Buchanan, New York 10511John D. O'TooleAssistant Vice PresidentConsolidated Edison Companyof New York, Inc.4 Irving PlaceNew York, New York 10003

Template:GL-Nav

Retrieved from "https://kanterella.com/w/index.php?title=NRC_Generic_Letter_1979-45&oldid=475102"