ML19322C678

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Technical Rept on Beznau Unit One Incident of 740820: TG-1 Trip/Reactor Trip/Safety Injection Actuation
ML19322C678
Person / Time
Site: Crane 
Issue date: 08/20/1974
From: Cecchi T, Galletly R, Lafaille J
WESTINGHOUSE ELECTRIC COMPANY, DIV OF CBS CORP.
To:
Shared Package
ML19322C677 List:
References
TASK-TF, TASK-TMR SA-251, NUDOCS 8001200002
Download: ML19322C678 (52)
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{{#Wiki_filter:__ o rs ( SA/251 TEC11!!ICAL REPORT ON BE2NAU UNIT ONE INCIDE';T CF AUGUST 20, 1974 : TG-1 TRIP / REAC'"CR TRIP /SAFITY !!! JECT!ON ACTUATION. i ... ~> / A j ^ j;, o J.P. LATAILL7 R. GALLETLY

I

~ T. CECCHI /1 ', 0 h. CO :<DLE, Direc cr, h Systems Engineerine i; Septe=her 2, 1974 7 80012000

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) TABLE OF CCNTENTS TSCHNICAL REPORT ON BE2NAU UNIT ONE INCIDENT OF AUGUST 20. 1974 : TG-1 TRIP / REACTOR TRIP / SAFETY INJECTION ACTUATION Pace I. INTRCDUCf!ON.. 1 II. SEQUENCE OF EVENTS 'l III. TPANSIENT BEHAVIOR OF.v.AIN PLANT VARIABLES 3 IV. DAMAGE TO THE PIPE RESTRAINTS A,ND SUPPORTS 5 V. EVALUATION OF THE INCIDENT 7 VI. OTHER RECC:L".ENOATIONS 14 VII. APPENDIX A 16 VIII. FIGURES (13) 20 l i e I I r e

1-I - IMTFCOUCTICt1 This report is produced as result of a site visit following the incident on Be:nau I which took place on August 20, 1974. The object of the visit was to make a rapid evaluation of whether the consecuences of the incident would jeopardice safety. This report, confirms the telex of Aug. 28, 74 on this subject. The secpe of this report, therefore,. is limited to a description of the secuence of events and of the damage observed, together with a possible explanation and assessment of safety issues. It is not meant to be a ccec.rehensive analvsis of the e.ffects of the incident. ~ . _.. ~, ~. II - SEOCENCE OF FVENTS CURING THE INCICEMT 1 Cn August 20, 1974, i a trip of one of the two turbines on the. Be=nau I reactor followef by failure of the steam dump system-, to operate resulted in a reacter trip and the opening of the pressuri:er relief valves. One of these valves subsequently-failed to close and the extended bicwdcun of the pressuri er resulted in the rupture of the prassuri:er relief tank, bursting disk. Examination following the incident revealed that the .pressuri:er relief valve which had failed to close had been ,' ; I. damaged, as had some of the supports to the pressurizer relief ' line itself. The secuence of events, with times where known, is recenstructed belev : Initial cenditiens Dawe : August 20, 1974 Tire : 11.20 a.m. Pressuri:er pressure

154 bar Pressuricer level : 501 Pressurizar relief tank le' vel
30%

Pcwer output of turbogenerator 1

137 "W (e) 2 : 177 MW (e)

I t

I ) Ti.me Event Disturbance occurs on the external grid network. TGl

  • trips out on high casing vibration.

~ 11 hrs 20 min 07.8 sec vibration causes low a p signal from hydrogen seal oil system. .L Steam dump valves fail to open. j SG steam pressures rise. Pressurizer pressure rises. Pressurizer level rises. 20 11.9 Both pressurizer relief valves open. 20 17..B - -Turbetrol of TG2 drops into the emergency mode. 20 23.0 one pressuri:er relief valve, closes in accordance with automatic, signal, pressure continues to fall and level' ) continues to rise. Pressurizer relief tank pressure rises. Pressuri er relief tank level rises. TG2 power level falls then rises to an overpower of 214 y.W (e). e 21' 00.4 4 Reactor trips on pressurizer icw pressure. 21 01.2 TG2 trips. SG steam pressures risa. SG water levels fall. Pressurizer level falls. 23 03.5 Secondary side safety valves lift. 23 13.9 Steam is fo=r.ed in the RCS het legs and pressurizer level rises past 100% and remains off-scale for 3 to 5 minutes. A reasonable assumptica is that water discharge occurs throuch the open relief valve. } Operator shuts pressuri:er relief line isolation valve. (Reperted vercally as 2 to 3 minutes af ter the trip). ../...

t 1 l Pressurizer level falls rapidly as steam bubbles in RCS collapse. Pressurizer relief tank bursting disk ruptures. Pressurizer relief tank pressure falls. Pressurizer relief' tank level falls, 11 hrs 23 min 43.5 see High containment pressure recorded (peak 1.1 ba abs.). 24 51.2 High containment temperature recorded (53.4'C). { 1 25 17.8 High. containment activity recorded ) 3 (17.3 mr/hr). i 32 14.J-' - SIS initiated as pressuri:er level falls l to St. Pressurizer level rises as SI water is .t... s added to t.ie RCS. u s SIS stopped manually. , Subsequently Procedure begun to bring reactor to cold shutdcwn condition usine the at=oc - phari: stecn rallof valves. Fig. 18 shows the record of pressuricer pressure and level transients followinc inciden- 'ad-d ation. i i '.. ~. i ,d e ,..I-s t

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.v.. --.2 .a-c.:n. w. ..n_.. -.. 3..,.C n.c a _ : vu.... .m_ A turbine trip in a two turbine plant is equivalent to a 50% load rejection and no reactor trip should be initiated if cen rcl syste=s work correctly. Since in Ee:nau the steam dump system did not work at all, initially the main variables behaved as follows :

1. Steam Generator steam pressure rese (to about 66 bars) but not enough in order to actuate safety valves.
2. Feedwater flow, steam flow and steam generatcr level decreased norma.3
v. as ex ected.

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3. The reactor being in automa' tic control, the nuclear pcwer decreased.

When reactor was tripped after about 49 seconds, it was at 761.

4. Pressurizer pressure rose rapidly frem 154 bars to a maximum of 160 bars (pressurizer relief valves actuation) in about 11 seconds.
5. Reactor coolant system averarje temperature rosc rapidly frcm 298.5*C to a maximum of 305.5'c in about 50 seconds
6. Cold leg temperature rose rapidly from 275*C to 290*C, then decreased to 240*C in 10 minutes, to 220*C.in next 100 minutes and 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 pressurizer relief valves were rapidly actuated. Their actuation ecok place almost simultaneously. However, it is very probable that the valve actuated by'the compensated pressure error signal (signal elaborated by a PID controller) cpened scme seconds before the other one due to the derivative term of the PID controller. When pressure decreased below relief valves actuation se: point the valve directi.v contrclied frcm an unce=r.ensated pressure' signal did not' shut.- This resul:ed in a depressuriza: en at rate of about 0.75 bar/sec, resulting in a reactor trip by icw pressur in approximately 49 seconds. The reac:c: trip signal tripped the turbine which was still in operatica, resulting in a further steam pressure increase (above 70 bars) which produced steam generator safety valves actua::cn, i lcwering the pressure :o about 65 bars. ) !l' ../...

y. I Reactor ecolant system average temperazure decreased to about ~ 285'C and pressuri:er level to 22% in about 1 minute after reactor trip. At this point pressuri:er pressure had fallen to hot leg saturation (70 bars). Subsecuently, hot leg flashing resulted in an increase of pressurizer level until the pressuri=er 1 filled about 3 minutes after reactor trip, resulting in probable licuid water discharge frem the relief valve and bulk boiling in the core. x Then the operator isoldted the failed relief valve, and pressurizar level decreased' reaching the setpoint (51) for safety injection actuation (safety injecticn is actuated by coincident low oressurizer pressure and level S.I. signals ) about 11 minutes after reactor trip. The system then started refilling. When pressurizer.. level reached about 70%, safety injection pumps were shut off =anually. The reactor was then' brought acr= ally to cold shutdown conditions. xy _ c~. u.nce-.n. u.. r. =..r. e r 2 or g.e..e.. 3. r. -e... g n-e., .e. t e. e. n..e. For pipe layout, see isc=etric, fig. I attached. The relief line to the pcwer relief valves comes cut of the 3 pressurizer : p and runs directly d wn (vertical run cf 6.8 m).' It passes through a grating flecr. No impac: evidence between the fico and the pipe ins,ulation exists. (Gap about 25 =m). At the bottom of the vertical run there is a conscle type restraint. (Locatien 1 in fig. 1). The main dirensiens are given in fig. 2. There is contact evidence, as shcwn en the figure, but no damage. The pipe then runs hori entally to the restr:in: 2 (fic. 1) This restraint limits =ction cf the pipe in a hori: ental direction. perpenficular to the pipe axis (See fig. 3). Scra ches en the shoes indica:e tha: he pipe coved ahcu: 26 .m axially. The sep' , cart cf the insulation is s11gh:1y seeshed (See fig. 3). l l ../... x ::uclear =cuer was : hen an os. The line then runs vertically down (2.77 m) and separates into two branches each having a step valve and a relief valve. ) Fig. 7, B and 9 show the damage to the valve. Examination of the pressuri:er relief valve which failed to close revealed that the yoke had broken off completely. One arm of the cast iron yoke had broken at the top and the other arm at the bottom taking part of the, yoke ring with it. The top break shcwed the presence of a very large flaw (inclusion). All broken faces shewed classic brittic failure together with evidence that the faces had rubbed together following failure. In addition it was reported 'that the valve spindle had been slightly bent. This was not observed since repairs had already. been started. g. Fig. 6 and 7 show the pedestal of the support between.the two valves. Fig. 4 is a sketch of the support and details the damage. The darage corresponds to a rotation of the pipe around a hori: ental axis perpendicular 'to the pipe axis. No evidence of translation has been found. Considering fig. 7, the back bolts were 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, but no other damage.could be found. After the valves the two branches of the pipe drop to the lower flocr. Fig. 10 shcws the penetration corresponding to the daraged branch. At the icwer ficor, the restraint R4 (See fig. 1) has been pulled off the floor (see detail in fig. 14). The =ctica has been imposed en the fra.me by the har of the hanger passing through a 50 mm slet in the frame (See fig. 11). ) ../... t% s

7-Pestraint RS, which is only a colu.~n supporting a sliding shoe, shows a me:ica of 70 mm as shown in f;g. 5. The pipe then joins a header and passes through the ficer (n6 en fig. 1). There is evidence of 25 mm upward displacement. At the icwer ficer the header has an elbow. Motion is restrained by a snubber. The bolts fixing the snubber to the concrete were found to be loose. V - EVALUATION OF THE IMCIOE!T"' This evaluation covers the incident transient effects and a preliminary esticate of magnitude and prehable causes 'of damage to the pressuri:er re. lief pinina'and sucports.

1. Cemoarisen with desien transients This Be:nau I incident is similar to the two fellcwing incidents which a:e nor= ally considered among reactor ceclant system design transients :

- Loss of Icad (up to pressurizer relief valves actua icn). - RCS depressuri:stien (frem pressuricer relief valves actuation). From the standpoints of core pcwer, heat transfers and syste=s pressures and temceratures, the reper ed incident is less severe than the design transients censidered above. The magni:ude and variation rate of the tempera:ure and pressure transients resulting from the inciden: are indeed fully ecvered by the values used for eculpment destgn. Plan variable behavier during the ::ansient did not resul: n an uncontrelled er damaging situa::en, and the released activ;;y ../...

o remained well below dangerous 11dits. All extsting protecti: l I systems (steam genera c sa'fety valves, reacecr trip, safety injection) worked properly and were adequate to handle One incident avoiding core and equipment damage.

2. Evaluation of damace to the cressurizer relief line, the relief valves and suceerts.

~ The relief line between the pressurizer and the power relief valves is part of the reactor coolant pressure boundary and thereftre is important to.the safety of the plant., 'The one po'. reg relief. valve which failed to close was isolated in accord with design intent by the operater closing the appropriate relief isolation valve and hence no uncontrolled loss of ecolant occurred. The review of the relief line ec.u1.= ment showed damage to the ) relief line supports and the pressurizer relief valve PCV-456. The damage evaluation and probable causes are treated below. a) Discu - - - - ssien of the incident- - - - - - - - - related to cause o f da.-ace. Examinaticn of the relief line and supports along with the records of pri.-ary Eeactor ecciant system parare:ers leads to the following cbservations. (1) It is probable tha: the observed damage to the supports is the resul: of hydraulic shecks from a sequence cf water and steam discharge through the relief line. (a) The pressuri:er relief line frcm the relief valve to the pressuri:er can fill w :h condensate. This } distance is approximately 19 meters, and can centain a volume of 0.06 m'. Cpening of the relief valves ../... t l l

~ will cause a rapid discharge of the water. The resulting dynamics are one possible cause of the piping displacements observed. (b) Based upon the recorder chart of pressurizer water level, it appears prehable that some water dischargc occurred later in the transient when the pr?ssuri:e: was conpletely filled.. The records indicate that this event could only have occurred aftar automatic closure of the undaraged valve (PCV-455C). Dynamics related to this event are another possible caus,e of the observed piping displacements and support damage. (2) It is not possible frem a'vailable evidence to provide one sequence of events which uniquely explains the cbserved results.cf the transient. It is not certain that.the valve damage was the consequence of the same hydraulic shock that.resulted in the support da= age. The cbserved sequence of events indicates that one a likely scenario is as follows (a) The undamaged relie f valve, PCV-4 55C, cpens firs: on the derivative compensated pressure centrcller a few seconds before the second valve cpens. (b) The wa:er slug fer ed by condensed pressuri:er steam in the relief line is largely discharged through the undacaged valve. We note tha: this pertion cf the line showed li::le er ne supper: damace. ../...

10 - ~ (c) The second valve, PCV-456, opens on continued pressure increase and the ::ansient, ecmbined with the large flav in the valve yoke results in valve failure. With this hypothesis, there is no reason to expect a hydraulic shock higher than in opening of the first valve hence pipi5g displacement sufficient to damage supports might not yc have occurred. (d) The first valve closes automatically upon a reducing pressure signal before pressurizer water level reaches 1001. i, (e) Water discharge occurs upon filling the pressuri:er creating a substantial hydraulic shock i$ the relief line. Since the undamag'ed valve has already closed the resultant pipe displacement was most pronou,ncec ' in the portion of line where the damaged valve is i located. i Other scenarios can also be postulated, but none has sufficient support of evidence to permit identificatien of a single secuence of events as the cause of observed damage. i. (3) The events which lead to complete filling cf the pressuri:er and the second water discharge through the relief line required more than a single failure (a) The failure of all the secondary steam dump valves to opera:e. Os) The failure of the rressuri:er relief valve := close.

-is likely that such a failure wculd ne:

../...

1A - ,o have occurred even with an initial hydraulic sheck, without existenc'e of a large flaw in the relief valve ycke. (4). Considering the valve PCV-4 56 itself, when in' the open position, there is a spring force p'roducing a tension of about 60,000 to 80,000 Newtons in the yoke. When the disk lifts, this force can be a plified due to dynamic effects. The presence of the flaw in one of the arms overstres. sed that arm (area reduction and stress concentration), which caused it to break. This enus d a mcment to be applied to the other arm, resultiva in bev&lec *of the spLndle and rupture. of the base.. The broken.'etal surface acpearance was typical of brittle failure with sc=e polishing due to rubbing contacts follcuing ycke separa:icn. The yckh ths rose about 2,5 cm, the nor=al stroke of the valve.- With the broken yoke, the valve failed to close. 1 Dynamic forces due to the free motien of the operator body =ay have contributed to damac.e to the su,c.c. ort. 4 i (5) Appendix A calculates the forces and stresses en the relief line piping in two locations, suspected to be among the =ost stressed. It is seen there that, within the calcula:1ca assumptien the piping could have been marginally overstressed. Ecwever, since a dye penetran: check of the PVC-456 valve to pipe weld was reper ed to show no defect, we cannet see any reascn to think that the. plant wculd cperate in unsafe condition with the line in the present s:2.t e. This sta:ement assumes of course tha; all the supper: sys:e= cf the piping will have been re:urned to its desien cond::icn befcre the reactor coes back to pcuer. ../...

To gain further assurance on the safety of the line we would recommend that a dye penetrant check of all welds near the fixed points be made a the earliest convenience. The locations include the pressuriser no::le, the relief tank no::le and the intermediate supported or restrained points. 4 b) Ocerationa - ~ - - - l Consideratiens (1) Plant operatien with one pressurizer pcwe- -=' d ef valve closed off dces not present 'a safety problem. The high' pressure reacec: trip and the pressurizer safety valves provide the necessary protecnicn against overpressure of the reacter coolant pressu're boundary. The existence of the power relief valves is to preven unnecessa.ry opening of the. main code safety valves I' i during certain plant design transients. e,- w~ (2) The safety injectic.n system functioned ner= ally with, a reporte,d ::tal in]ection rate of 4 0 1/sec. The injected water raised the pressuri:er level from ST to-751. Assuming the injection water to be initially at 16*C and at=cspheric pressure in the RWST and Oc end up in the pressurizer at 285'C and 110 bars then the quantity of water leaving the RMST must have been abou: 3 10 m. This would cause a decrease in RFST level of about 0.7%. The injection time would he abou: 4.1/2 minutes assuming a constant in]ection rate. ) l /

- 13 (3) The reason why the turbetrol gear of turbine 2 dr=pped into the emergency mode is not known.

was reported that the affeet of this would be to lock the turbine inlet control valves in their last position.

Thus they would no longer respond to changes in steam pressure. This may account for the overpower excursion experience on turbogenerator 2 just prior to its tripping. (4) The failure of the staan dump valves to open was reported to be che, result of a wrong viring connecticn which was no: discovered during' testing. The control circuitrv of the steam du.p valves had been out,for maintenance at sene previous da:e. Before being put back on line, the circuitry had been' ested in two halves. F.ach half was checked independen:1v. of the other half and each half' checked out satisfactorily. A fault at the interface of the two halves thus rerained unrevealed. g 9 I e l g 3 3 p 6 e t o j

vt - CTHI? RIcouMI:: CAT:or:s )

1. The piping displacements and support damage which occurred have indicated ~the possibility that the pressurizer relief line was narginally overstressed.

The likelihcod is that the displacements resulted from either discharge of a water slue initially in the line or from relief of wa:er when the ' pressurizer was coepletely filled. The initial evaluation o' stress was deduced frc= ebserv support displacerent and support helt strains. As such, no , definitive indication of pcssible stress levels with this transient exists as basis fer an evalua icn of fatigue damage . for the entire piping length. We would reco==end a dynamic analysis be performed , consideri-at a mini =um the effects of the stear condensate initially ,) in the line. The force ti.T.e history function can then be used for evaluaticn of fatigue damage as well as the adecuacy of restraints.

2. The failure of the power relic f valve yoke is =cre probabl due to the use of cast-iron materials of ccnstructicn wh e

i= pac: ere prcperties are poor and flaws of the type invcived in this failure can remain undiscovered. We therefere recc= rend such non-destructive tests as a feasible be made :o ascertain that re in the valve currently installed. no flaws of this type exist Further considera:icn mich: with a less brittle material.he given Oc replacing these yokes ~ ~ ) ../... i l l

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3. ':'he tes: procedures following maintenance of the control system to the steam dump v51ves should be rewr:etten to eliminate the possibility of unrevealed faults.

f 4 It would be useful to provide means (i.e. 2 separate alarms : one actuated by the unccmpensated pressure signal and the e other bv the comnensated nressure error signal) in order to know if certain1v each pressuri=e'r relief valve opens during a pressure excursion. e e f S I e e e e e e e 9 t e 9 e s 80 ,g e e G 9 g e s =, g I %^ 1 e e 6 e ae s e,. e e e e ,j ' ? i 8 ,8 g e i e 0 4 0 lee e e e e e

gg A-1 APPE?CIX A ) Stress and Toree Evaluatten in the nice between valves _1_. _.D. _a r_ _e c_ _e _ _t _o_ _t _h e__s _u_n_c_e_r_t The two bolts on the right side on figure 3 were strained about 3 mm. The two bolts on the left side were also strained but only to the point of getting loose. 4

2. E

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_o m_e.n._t_ _a _a_n_l _i _e _d_ _t.o._ _t _h _e_ _s _u _n _ne_ _r_t 1 i Bolt site : M10 + Shaft si:e* (diameter) 8.888 < d < 9.128 =m (Catalogue MARC-GIRARD - 1970) 3,, Section (average) 3 (8.888

  • 9.12s) 2 2-63.73 =m

= 4 2 Assume for the bolt material a yield stress of = 32 kg/:r..2 o 1 y Hence the me=ent to strain the two bolts is i M = 63.73x22x2x.125 = 550.6 kg.m I i g e 2a_ESI55_ISE' die 2$_;o_ggsa;g_g.3;_ggggg; i J "o 8 e f I 1 R2 F 385 R, I t 4 8 ep .;50 4C5 3GC j A-2 If one norlects the effect of the supports located downstream of valve 456, one can write ahe equation 385xP = 135xR L Knowing that R x.135 = 550.6 kgm g Hence F = 1430 kg It is felt that such a force is in the possible range. da_EEIf 8295_iU_ IDS _DID3 (Prima ry stresse only) Pipe : 3" sch 160 !!ence : 00 = 3.5 in = 88.9 mm t= 11.13.=m Bending modulus = A' = 47.17 10 mm 3 3 v i Bending stress : b 250.6 10 = 11.67 kg/mm e B I/v 47.17 104 Pressure stress (ASME III, Article ::B 36 52) -2 exOD 164.5e10 x83.9 C 6.5, ,c/== o= = = r. 2t 2x 11.1 ;- Combination (Article !!D 36 52) PD. D +P D. tt 1 2: 2 21 i S and 3 are taken frcm ta le 3683.2-1 g 2 B =3 y 2" I ~ N'"C" 6.57 + 11.67 = 1S.24 kg/m? = (, g

'** A-3 _5_. _ A_ _l _lowa _b l e _ _s_t_re _s _s _e _s SA 376 Grad 2 316 2 S at room tume. = 20 ksi = 14 kc/mm m 5,at 650 *F '(=34 3*C) = 16.6 ksi = 11.6 kg/mm Allowable stress = 1.5 S, (ASHE III, article ND 36 52) 2 1.5 S, = 21 kg/en (room temperature) = 17.4 hg/mm* (343*C) 6. --__C_o_n_c_l _u _s _i o_ n f _o_r__c_r_i.m_a rv___s t_re _s _s _e _s _ _i n _t_h e _c_in e Since it appears that hot fluid has been carried by the pipe for a time of about 3 min, the hot al,lowable stress needs to be taken. Then it appears' that the actual stress is slichtly higher than the allowable : / 2 18.24 > 17.4 kg/cm o It should be noted that the figure of 13.24 k;/mm? is a mini. mum, since it corresponds to the plastification of the support (M = 550. 6 kge). _7. __P r_ima _r" a n_ d_ S_e_c_o_nd_a_rv_ s t r_e_s_s_e_s_ _i_n__t_he c _i_c e The evaluation of secondary stresses (art'icle N3 3653.1) rneuires the knowledge of the temperature gradients in the pipe. It was thus not pessibic to evaluate these stressos. E EilCGE2_s5gggsgs_g5_3bg_;gdygg; Bending ec. ment ( 3 35 - f (4 05 - 13 5)) k g mm M= 1430x 357 kejr =

15 - A-4 reducer 21 " sch 16C CD = 2.875 in = 73.02 mm t= .375 in = 9.52 mm I 3 3 - =,1.64 in 26.9 cm = v 2 l - Pressure stress = oxOD = 6.28 kn/mm 2,. t 2 Bendina stress =.'/v = 13.28 kg/cm 1 2 fotal stress = 19.56 kg/mm This stress should be considered more as indicative s nce it depends so much on the assumption of the force location. The sane conclusion holds as for the pipe stress. B e e

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NOK REPORT ON BE7.MAU ACCIDC;T OF AUGUST 20, 1974 1. TRI? TG-1/ REACTOR TRIP /SI/ on Aucust 20, 1974 at 11:20 a.m. a trio on turbine TG-1 occurred resultinc to hich bearing and casinc vibrations (Bearinc 6:60 ) At trio time, generator 2 was deliverinc about 140 MVar. Resulting from a failure of the steam.duno system to operate, with the consecuence tha: the relief valve did not open. That resulted in a racid rise of coolant temperature, steam cressure and pressuri:er level and pressure. At 160 har of pressure in the crimary, the =ressuri=er cressur-relief valves ocened', lowerine ranidiv the cressure in t*s crimary. About 10 seconds after valve c=eninc, the cressure had reached such a lcw level that the cressur-1:er pressure relief valves were reactuated to close. Due to a disturbance, valve PCV-456, failed te close, resultine in a lowerinc of RCS cressure uo te 100 bar after about 1 minute. Reactor trigned resultinc from a lcw pressure sicnal (126.5 bar). Due to the c=eninc of the pressuricer relief valve, the pressure in RCS drop ed to about 70 bar, correscendine to a saturation tem =erature of 234*C. Consecuentiv, staan acceared in the prinary het lec, filline the pressuri er. Two or 3 minutes after tric, the ccerator recocnised the failure of the relief valve and tsela ed it with the ecwer l l o=erated valve 531. The water level becan to dren, and 11 ninutes after trin, autenatic S was initiated hv Icw r l cressure and level in :: e cressurizer. 1 l [

Pace 2. ) SI systems worked normallv and about 40 litres per second of water was spilled throuch the four SI oumo nozzles into the primary, causing a rise of pressure to llo bars and a further rise of level to 70 %. The SI pumos were then turned off and the cower operated valves of the scray pioings were closed. From that moment on, the cressuricor level could be controlle throuch charcinc pungs and release of steam, assunina the .,' orinary to cool down. About 3 minutes after trio, the containment oressure alarm signal was actuated because of too high cressure, and 1 minute later the hich activity alarm. Maximum cressure in containment reached 100 mbar over nornal. The operators activated the containment fan coolers. Since several safety alarms of the cressuri=er relief tank were on, it was cuickly assuned that the ructure disc was broken and that the discharce channel was defectuous. After TG-1 trio, due to steam dumn failure, steam cressure rose to 66 bar. The turbetrol of TG-2 was actuated as an emercency after TG-1 trio. TG-2 was unregular in behaviour, and the position of the control valve remained constant during the cressure transient. The cerfornances of TG-2 rose to about 214 MWe due to hicher steam cressure (rise from 52 bar to 66 bar). After TG-2 trio, followine reactor trio, steam cressure rose to over 70 bar, actuatinn the safe:v valves and thus lowerinc cressure to about 65 bar. ) 2'. CUROMCLOG* CAL 5? Fur?TE OF r7-NTS Aucust 20, 1974

PEga 3. 2.1. Reactor Trio Becinninc of incident 11 h 20' 12" TG-1 main breaker off Pressuri=er nressure low-trip 39,7" later ~ Reactor trio breaker oeen 39,G" later TG-2 main breaker off 40,3" later SI actuation (pressurizer pressure and level low) 11'55,3" later 2.2. Events as Recistered on Alarn Tveewriter TIME 11:15 TG-1 power hich 135,5 MVar 11:20 Allowable oil

essure of TG-1 too low 11:20 Pressuri=er prassure 158.2 har high.

11:. 20 Pressuri=er =ressere 159.9 bar high. Reactor Trip. 11:21 Tave RCS-A high 302.2*C 11:21 Stean nr. uestrean of 66.3 bar TG-1 stoo valve hich, 11:21 Tave RCS-A hich 305.2*C 11:21 SG-A stean cressure 67.3 hsr hich. 11:21 SG-R steam cressure 67.2 bar

high, 11:21 Steam =r. unstrean of 77.6 har TG-1 sece valve.

11:21 SG-A stean =ressure 73.3 bar hich. 11:21 SG-A steam cressure 65.4 bar j hich. ( 11:22 Safety oil nressure of l TG-2 000 lcw. l 11:22 Tavg RCS-A 285.2*C

Pcqo 4. ) TIME ~ 11:23 Steam pressure umstrean of 68.1 bar ~~ TG-2 stco valve. 11:23 Pressurizer relief tank 62.8'C temperature hich. 11:24 Pressurizer level 79 % 11:24 Pressurizer level 88 % 11:24 containment oressure high 1.1 bar abs. 11:24 Pressurizer. relief tank level 20.2 % low. 11:24 Pressurizer relief tank cressure 0.59 bar high. 11:25 Pressurizer relief tank cressure 0.15 bar 11:25 SG-A*B steam cressures normal. 63.7 bar 11:25 Containment activity hich 17.3 mr/h 11:26 Loop B RCS flow low. 38 % 11:27 Containment air tencerdture hich 53.4

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6.8 % 11:32 Pressurizer level normal. 18 % 11:33 Surge line tencerature too low. 271.l*C 11:34 Pressurizer levelshich. 58 % 2.3. Secuence of Events for Pressurizer and Pressurizer Relief Tank TIME 11 h 20' 11.1" Pressuri er cressure above control rance. 11.9" Pressuri:er relief valve. 22.8" Pressurizer relief tank cressure hich 23.0" Pressuri er relief valve icoked 23.0" Pressurizer cressdre nornal 23.1" Pressurizer relief :ank level hich 24.2" Pressuricer level hich. 33.0" Pressurizer relief tank cressure too hich. 35.0" Pressuri:er cressure under nornal. Ii i l

Paco 5. TIME 11 h 21' 00.4" Pressurizer cressure low - Trio. 01.2" Pressuri:er cressure low - SIS unlocked. 05.1" Pressurizer relief tank level hich. 13.5" Pressurizer pressure low - SIS unlocked. 11 h 23' 27.6" Pressurizer level hich - 1 channel tri-43.3" Pressurizer relief tank level too high 43.5" Containment oressure too high, 47.1" Pressurizer relief tank level icw. 11 h 24' 29.4" Pressurizer relief tank cressure norna. 51.2 Containment temperature hich. 11 h 25' 17.8" Containment activity hich.

3. ANALYSIS OF PFE CAUSES OF TFE INCID7NT TG-1 trioned due to hich casing vihrations, esoecially in casing 6.

It had already been noticed that TG-1 was sensi~tive to shocks. At the menent of incident, TG-1 was set to function under maximum effort, se that it could supcort a naxinum of vibrations. r The tric is not unfamiliar and would not have affected the primary if steam dumo had normally been actuated. i An inspection of containnent after orinary had ccoled down, showed that the yoke between the :CV-4 56 valve housine and air enpine was broken, and prehably due to a dynamic effer: en the pipine at cpening of the valye. Consecuently, the valve failed to close and initiated a ra=id fall o' cressure in erinary. The pressuriner relief tank ruuture disc breke, due to a Orclenced sur:e of crinarf ecciant in the tank. Itens 2 and 3 shew the disc hrcke when the relief valve had already closed.

Paqa 6 WATF.R COLLF.CT"O IN CONTAINMENT SUMP 3 Recen. hold up water Tank A 38 % - 100 % = 9.8 m 3 Regen. hold up water Tank 3 16 % - 36 % = 3.2 m 3 Total quanlity of water collected n13.0 m Pressurizer relief tank 80 % - 19 % =11.2 m Water out of system. = 1.8 m Since no further damace was noticed in containment, it could be assumed these 1.8 m of water were blown out. 4.1. Thermal Stresses in RCS Beside a rapid water tenperature rise of abcut 6*C after TG-1 tripped, a racid prinary cressure rise from 154 har to 160 bar, there was also an incertant temperature transient in area of SI no::les. However, since the ) reactor's main pumps o=erated all the time, thus mixinc' cold snray water with hot coolant, it can he assumed that other components didn't under o hich te.meerature cradients. Furthernere, no::le temperature and stress remained within desian limits. 4.2. Damaces to Relief Svstens Durine insnaction in containment after coolinc cf crimary, the followinc dacaces in the cressurizer relief systems were observed : - relief valve PLV 456 : Mechanis, broken on both sides and her

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- One ancher ecint of the relie' syste= nininn after valve - Relle: cahk =ressure disc broken. was loose ~. ) / Further daraces in centainment were not ncticed.

Paca 7 It must be said that the relief tank is not desicned to acceot stean from the cressuri:er for a crolenced sine. The danaces to the relie.# valve is therefore a direct cause to the breaking of the ructure disc. 4.3.

  • urbines TG-1 The cause of vibrations to the casine are most nrobably the stresses and shocks.

The P sienal from hydrogen seal oil system is due to casine vibrations. Damages to the seal or casing are nest im=robable. TG-2 The oscillation from 172 MWe to llo MWe, and then to 215 MWe suggested that the bolts of the high pressure cylinder were loosened and had lost some ofitheir tension. A too small stress was noticad, due to leakage of the seals of the high pressure cylinder. Due to too hich rotational mcmentum at 215 MWe, the coucline between turbine and generator was closely controlled. 5. When reviewine the secuence of events, the failure of two syst2ms, nanely the steam duno and the cressuri:er relief system, we came to the conclusien that it did not brine to an uncontrolable nor a damacine situation. Durine the incident, no activity (in cas or licuid fern) in the surroundinc area reached an uncontrollable level. The generater safety valves maintained the stean cressure within alicwable linits. "he S S broucht back the crinary to a safer cressure, allowine normal ecoldown conditions. 6. .,s v. u s. , -,,, D.. a..,-.c s s r % evn.. 6.1 Centrcl c' cenerator 1 Generator 1 reaching ranidiv to casine vibrations,i: will

Pena 8 e be tried to see if the reculator can be modified in order to have a quick action. 6.2. Pressure Reculator Tests will be made to see if the first row of imcellers in the cressure reculator of the turbine must not be reviewed i in order to limit power to 190 MWe. 6.3. Stean Dune Svsten a) Revisions and calibrations should be made in stean duno system (before o=ening of steam duno valve.) [ b) Studies will be made, to.ake periodic controls of steam dure while in o=eration. It should hele to insure better safety linits (for exanole : unwanted onening of steam dump valve). c) A control tyre writer linked to the steam dune will he installed in' order to control the opening of steam dunc valves and to check the cood workinc of oil me==s. 6.4. Pressuri:er Relief Svsten The first measure to be taken, is to recair the damaged ~ valve, the picine su== orts and review bcitirns. 2 The cressuri:er relief tank ru=ture disc must be reclaced. With these re airs star:-uc shcule> be nossible. To see how the relief systen nicine can be better secured and hcw shock at opening of relief valve can be avoided are further neasures to he taken. l t

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