NRC Generic Letter 1979-45
| 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, Three Mile Island, 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  |
| Issue date: | 09/25/1979 |
| From: | Ross D Office of Nuclear Reactor Regulation |
| To: | |
| References | |
| NUDOCS 7911260004, GL-79-045 | |
| Download: ML031320181 (71) | |
- ' ti- rth EGas tw UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON D. C 20555 Sb -3As- SEP 2 5 1979 TO ALL POWER REACTOR LICENSEES
SUBJECT: TRANSMITTAL OF REPORTS REGARDING FOREIGN REACTOR OPERATING
EXPERIENCES
The enclosed reports are provided to you for information and use in your reactor evaluations in light of the Three Mile Island Unit 2 accident. Enclosure 1 is an internal Westinghouse report which describes an incident involving a stuck-open power-operated relief valve that occurred 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's Special Commission investigating the TMI-2 accident. Enclosure 2 is an internal NRC staff memo on this incident. Enclosure 3 is a report on a steam generator tube "rupture" incident at the Doel 2 nuclear power plant in Belgium.
If you have any questions about the enclosed information, please let us know.
D. k'Ross, Jr., Director Bulletins and Orders Task Force Enclosures:
1. Technical Report on Beznau Unit 1 Incident of August 20, 1974:
TG-l Trip/Reactor Trip/Safety Injection Actuation
2. Memorandum dated May 15, 1979; Ashok Thadani to D. F. Ross, Jr.
3. Memorandum dated September 13, 1979;
Darrell G. Eisenhut to Multiple
Addressees
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1,4/
q 1)lg(lOovcf 640
. i -4 j To : O.A. Wilson (with att.) : T. Cecchi
(3 copies) Date : tSeptember 4, 197.
cc : F. Noon (with att.) Ref : SA/251 H. 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, 1974 References (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 Incident of August 20, 1974 prepared by WNE inspection team who went to Beznau on August 23.
This report, which should be sent to Beznau, summarizes our observations on the course of the transient, the damage as we viewed it, our calculations and conclusions.
Despite what is indicated in the referenced (2) letter, in order to have a more complete report, we added some recommend- at.ons for future changes. / T-
T. CrrC-CT!
- < SYSTEMS ANALYVS
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a%- / \Gu SA/251 TECHNICAL REPORT ON BEZNAU UNIT ONE
INCIDENT OF AUGUST 20, 1974 TG-1 TRIP/
REACTOR TRIP/SAFETY INJECTION ACTUATION.
J.P. LAFAILL
R. GALLETLY
T. CECCHI ,
H. CORDLE, Director, Svstems Fnnineerina September 2, 1974
DISTRIBUTION
H. CORDLE
A. HALL
D. ten WOLDE
0. WILSON
L.--BARSHAW
T. CURRIE
R. GALLETLY
F. NOON
J. LAFAILLE
T. CECCHI
R. LEHR
J. MCADOO
R. CLOUD
W. ROCKENHAUSER
TABLE OF CONTENTS
TECHNICAL REPORT ON BEZNAU UNIT ONE INCIDENT OF
AUGUST 20, i974 : TG-1 TRIP/REACTOR TRIP/SAFETY
INJECTION ACTUATION
Pace I. INTRODUCTION., - 1 II. SEQUENCE OF EVENTS 1 III. TRANSIENT BEHAVIOR OF MAIN PLANT VARIABLES 3 IV. DAMAGE TO THE PIPE RESTRAINTS AND SUPPORTS 5.
V. EVALUATION OF THE INCIDENT 7 VI. OTHER RECOMMENDATIONS 14 VII. APPENDIX A 16 VIII. FIGURES (18) 20
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-- l1 T - INTRODUCTION
This report is produced as a result of a site visit following the incident on Beznau I which took place on August 20, 1974.
The object of the visit was to make a rapid evaluation of whether 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 description of the sequence of events and of the damage observed together with a Dossible explanation and assessment of safety issues.
It is not meant to be a corprehensive analysis of the effects of the incident.
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II - SEOUENCE OF EVENTS DURING THE INCIDENT
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On Aucust 20, 1974, a trip of one of the two turbines on the Beznau I reactor followed by failure of the steam dump system,.
to operate resulted in a reactor trip and the opening of the pressurizer relief valves. One of these valves subsequently.
failed to close and the extended blowdown of the pressurizer resulted in the rupture of the pressurizer relief tank,burstina disk.' Exarnination following the incident revealed that the pressurizer relief valve which had failed to close had been damaged, as had some of the supports to the pressurizer relief '
line itself.
The sequence of events, with times where known, is reconstructed below :
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)
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Time Event Disturbance occurs on the external grid network.
TG1 trips out on high casing vibration.
11 *hrs 20 min 07.8 sac Vibration causes low A p signal from hydrogen 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 emergency mode.
20 23.0 One pressurizer relief valve closes in accordance with automatic signal, . .. .. I
pressure continues to fall and level continues to rise.
Pressurizer relief tank pressure rises.
Pressurizer relief tank level rises.
TG2 power level falls then rises to an overpower of 214 MW (e).
21 00.4 Reactor trips on pressurizer low pressure.
21 01.2 TG2 trips.
SG steam pressures rise. I
SG 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 and pressurizer level rises past 100% and remains off-scale for 3 to 5 minutes.
A reasonable assumption is that water discharge occurs through the open relief valve.
Operator shuts pressurizer relief line isolation valve. (Reported verbally as
2 to 3 minutes after the trip).
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-3- 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 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 falls to 5%. L
Pressurizer level rises as SI water is added to the RCS.
SIS stopped manually.
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Subsecuently Procedure begun to bring reactor to cold shutdown condition using the atmos- phe:4o steam reliaf valves.
Fig.. 18 shows the record of pressurizer pressure an d level transients following incident initiation. * . *. *- .
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III - TPANSIENT BEHAVIOR OF MAIN PLANT VARIABLES DURING TEE INCIDES
A turbine trip in a two turbine plant is equivalent to a 50% load rejection and no reactor trip should be initiated if control systems work correctly. Since in Beznau I the steam dump system did not work at all, initially the main variables behaved as follows :
1. Steam Generator steam pressure rose (to about 66 bars) but not enough in order to actuate safety valves.
2. Feedwater flow, stebm flow and steam generator level decreased normally as expected.
3. The reactor being in automatic control, the nuclear power decreased. When reactor was tripped after about 49 seconds, it was at 76%.
4. Pressurizer pressure rose rapidly from 154 bars to a maximum of 160 bars (pressurizer relief valves actuation) in about
11 seconds.
5. Reactor coolant system average tcmpcraturc rose rapidly froin
298.5iC to a maximum of 305.5*C in about 50 seconds,
6. -Cold leg temperature rose rapidly from 275°C to 2906C, then decreased to 240'C in 10 minutes, to 2200C.in next 100 minute 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 took place almost simultaneously. However, it is very probable that the valve actuated by the compensated pressure error signal (signal elaborated by a PID controller) opened some seconds before the other one due to the derivative term of the PID controller.
When pressure decreased below relief valves actuation setpoint the valve directly controlled from an uncompensated pressure'
signal did not shut. This resulted in a depressurization at ratE
of 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 in operation, resulting in a further steam pressure increase (above
70 bars) which produced steam generator safety valves actuation, lowerinc -he pressure to about 65 bars.
. . /. . .
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Reactor coolant system average temperature decreased to about
285'C and pressurizer level to 23% in about 1 minute after reactor trip. At this point pressurizer pressure had fallen to hot leg saturation (70 bars). Subsequently, hot leg flashing resulted in an increase of pressurizer level until the pressurizer filled about 3 minutes after reactor trip, resulting in probable liquid water'discharge from the relief valve and bulk boiling valve, in the core. x Then the operator isolated the failed relief for and pressurizer level decreased reaching the setpoint (5%)
safety injection actuation (safety injection is actuated by about coincident low nressurizer pressure and level S.I. signals)
11 minutes after reactor trip. The system then started refilling.
When pressurizer-level reached about 70%, safety injection pumps were shut off manually.
The reactor was then brought normally to cold shutdown conditions.
IV - DAAGE TO TH.E RELIEF PIPE RESTRAINTS AND SUPPORTS
For pipe layout, see isometric, fig. 1 attached.
the The relief line to the power relief valves comes out of m).'
pressurizer top and runs directly down (vertical run of 6.8 It passes through a grating floor. No impact evidence between the floor and the pipe insulation exists. (Gap about 25 mM).
At the bottom of the vertical run there is a console type restraint. (Location 1 in fig. 1). The main dimensions are given in fig. 2. There is contact evidence, as shown on the ficure, 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 the shoes ind.cCate that the pipe moved about 26 FM m axiay. The top part of the insulation is slightly swishes (See fig. 3).
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3:Nuclear Pcwer was to.--. P. -..
I -. ,
The line then runs vertically down (2.77 m) and separates into two 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 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 voke ring with it. The top break showed the presence of a very large flaw (inclusion).
All broken faces showed classic brittle 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.
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 damage corresponds to a rotation of the pipe around a horizontal axis perpendicular to the pipe axis. No evidence of translation has been found. Considering fig. 7, the back bol _s 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 floor. Fig. 10 shows the penetration corresponding to the damaged branch.
At the lower floor, the restraint R4 (See fig. 1) has been pulled off the floor (see detail in fin. 14). The motion has been imposed on the frame by the bar of the hanger passing through a 50 mm slot in the frame (See fic. 11).
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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 restrained by a snubber. The bolts fixing the snubber to the concrete
'*'ere found to be loose.
V - EV.AwLUATION OF THE I!CIDENIT
ThiS evaluation covers the incident transient effects and a preliminarv estimate of magnitude and probable causes of damage to the pressurizer relief vinina and supports.
1. Comvarison with desian transients This Beznau I incident is similar to the two following incident which are normally considered among readtor coolant system design transients :
- Loss of load (up to pressurizer relief valves actuation).
- RCS depressurization (from Pressurizer relief valves actuation).
- *
From the standpoints of core power, heat transfers and systems pressures and temperatures, the reported incident is less sever'
than the desicn transients considered above.
The magnitude and variation rate of the temperature and pressure transients resulting from the incident are indeed fully covered bv the values used for equipment design.
Plant variable behavior durina the transient did not result in an uncontrolled or damaging si:uation, and the released activity
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remained well below dangerous lim~its. All existing protection systems (steam generator safety valves, reactor triD, safety injection) worked properly and were adequate to handle the incident avoiding core and equipment damage.
2. Evaluation of damace to the Pressurizer relief line, the relief valves and suonoorts.
The relief line between the pressurizer and the power relief valves is part of the reactor coolant pressure boundary and therefore is important to.the safety of the plant.
The one poster relief .valve which failed to close was isolated in accord with design intent by the operatcr closing the appropriate relief isolation valve and hence no uncontrolled loss of coolant occurred.
The review of the relief line equipment showed damage to the relief line supports and the pressurizer relief valve PCV-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 the records of primary reactor coolant system parameters leads to the following observations.
(1) It is probable that the observed damage to the sunports is the result of hydraulic shocks from a sequence of water and steam discharge through the relief line.
(a) The pressurizer relief line from t;.e relief valve to the pressurizer can fill with condensate. :his distance is apprcxiratelv 19 meters, and can conwain a voluzne of 0.06 m'. Openinc of the relief valves
. . /. .
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 probable that some water discharg;
occurred later in the transient when the pressurize:
was completely filled. . The records indicate that this event could only have occurred after automatic closure of the undamaged valve (PCV-455C).
Dynamics related to this event are another possible cause of the observed piping displacements and support damage.
(2) It is not possible from available evidence to provide one sequence of events which uniquely explains the observed results of the transient.
It is not certain that the valve damage was the consequence of the same hydraulic shock that resulted in the support.damage.
The observed sequence of events indicates that one likely scenario is as follows :
(a) The undamaged relief valve, PCV-455C, opens first on the derivative compensated pressure controller a few seconds before the second valve opens.
(b) The water slug formed by condensed pressurizer steam in the relief line is largely discharged through the undamaged valve. We note that this portion of the line sahowed little or no su=mort damage.
I , Ir (c) The second valve, PCV-456, opens on continued pressure increase and the transient, combined with the large flaw 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 pipizg displacement sufficient to damage supports miaht not yet have occurred.
(d) The first valve closes automatically upon a reducinc pressure signal before pressurizer water level reaches 100%.
(e) Water discharge occurs upon filling the pressurizer creating a substantial hydraulic shock in the relief line. Since the undamaged valve has already closed, the resultant pipe displacement was most pronounced in the portion of line where the damaged valve is located.
Other scenarios can also be postulated, but none has sufficient support of evidence to permit identification of a single sequence of events as the cause of observed damage.
(3) The events which lead to corpleze filling of the pressurizer and the second water discharge throuch the relief line required more than a single failure :
(a) The failure of all the secondary steam dump valves to overate.
(b) The failure of the pressurizer relief valve to close. It is likely that such a failure would not
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have occurred even with an initial hydraulic shock without existence of a larqe flaw in the relief valve yoke.
(4) Considerina the valve PCV-456 itself, when in the open position, there is a spring force producing a tension of about 60,000 to 80,000 xewtons in the yoke. W-hen the disk lifts, this force can be anplified due to dynamic effects. The presence of the flaw in one of the arms overstressed that arm (area reduction and stress 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 was typical 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 operator body 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 be among the most stressed. It is seen there that, within the calculation assumption the piping could have been marainally overstressed. However, since a dye penet-anm check of the PVC-456 valve to pipe weld was reported to show no defect, we cannot see any reason to think that the plant would operate in unsafe condition with.
the line in the present sta.te. This statemen- assumes of course that all the support sxystem Of t-.e piping will have been returned to its design condition before the reactor goes back to pcwer.
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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 at the earliest convenience. The locations include the pressurizer nozzle, the relief tank nozzle and the intermediate supported or restrained points.
b) Ooerational Considerations
(1) Plant operation with one pressurizer power relief valve closed off does not present a safety problem. The high pressure reactor trip and the pressurizer safety valves provide the necessary protection against overpressure of the reactor coolant pressure boundary.
The-existence of the power relief valves is to prevent unnecessary opening of the main code safety valves during certain plant design transients.
(2) The safety injection system functioned normally with, a reported total injection rate of 40 1/sec. The injected water raised the pressurizer level from 5% to
75%. Assuming the injection water to be initially at
16'C and atmospheric pressure in the RWIST and to end up in the pressurizer at 285°C and 110 bars then the quantity of water leaving the RTHIST must have been about
10 M 3 . This would cause a decrease in ?WST level of about 0.7%. The injection time would be about 4.1/2 minutes assuring a constant injection rate.
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2 dropped
(3) The reason why the turbotrol gear of turbine into the emergency mode is not known. It was reported that the effect of this would be to lock the turbine the- inlet control valves in their last position. Thus would no longer respond to changes in steam pressure.
This pay account for the overpower excursion experience on turbogenerator 2 just prior to its tripping.
! (4) The failure of the steam duzrp valves to open was reported to be the result of a wrong wiring connection
! wh-ch was not -iscovered during testing. The control circuitry of the steam dump valves had been out for maintenance at some previous date. Before being put back on line, the circuitry had been tested in two halves. Each half was checked independentlv'of the other half an6 each half checked out satisfactorily.
A fault at the interface of the two halves thus remained ur.revealed.
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1' o VI - oTHER RECOlMMENDATION:S
1. The piping displacements and support damage which occurred have indicated the possibilitv that the Pressurizer relief line was marginally overstressed. The likelihood is that the displacements resulted from either discharge of a water slug initially in the line or from relief of water when the pressurizer was corpletely filled.
The initial evaluation of stress was deduced from observed support displacement and support bolt strains. As such, no definitive indication of possible stress levels with this transient exists as basis for ad~evaluation of fatigue damage for the entire piping length.
We would recommend a dynamic analysis be performed, consideri:
at a minimum the effects of the steam condensate initially in the line. The force time history function can then be used for evaluation of fatigue damage as well as the adequacy of restraints.
2. The failure of the power relief valve yoke is more probable due to the use of cS~t-- onmateriads of Q~sruction where impact Properties are poor and flaws of the type involved in this failure can remain undiscovered.
We therefore recommend such non-destructive tests as are feasible be made to ascertain that no flaws of this type exist in the valve currently installed.
Further consideration might be given to replacing these yokes with a less brittle material.
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3. The test procedures followino maintenance of the control system to the steam dump valves should be rewritten to eliminate the possibility of unrevealed faults.
4. It would be useful to provide means (i.e. 2 separate alarms one actuated bv the uncompensated pressure signal and the other bv the compensated nressure error sional) in order to know if certainly each pressurizer relief valve opens durina a pressure excursion.
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APPENDOX A
Stress and Force Evaluation ih the nine between valves
1. Darace to the sunoort The two bolts on the right sidce 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.
2. Evaluation of the moment aonlied to the sunnort Bolt 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 of a 32 ka/mm2 Hence the moment to strain the two bolts is M - 63.73x32x2x.135 - 550.6 kg.m
3. orce -ecu4red to create that moment
. ~.
fl ~38;5 t T
lVA L *EP ?C1 456
5LAL 33/
1460 405 _ _ 00
i
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- 17 - A-2 If one neclects the effect of. the supports located downstream of valve 456, one can write the ecuation
385.F = 135%R.1 Knowino that R x.135 = 550.6 kgm Hence F = 1430 ka It is felt that such a force is in the possible ranne.
4. Stresses in the nip2 (Primary stresses only)
Pipe : 3" sch 160
Hence : OD = 3.5 in = 88.9 mm t = 11.13 mm Bending modulus = v= 47.17 10 mm3 Bendina stress :
- ~ 32 a = M 550.6 10 = 11.67 kg/mm2 B Pe/r s s 47.17 1t Pressure stress (ASI'E III, Article NB 36 52)
ap = vOD -
164.5x10 2 .88 .9 6.57 kq/mm
2x 11.13 Combination (Article NB 36 52)
P1D + 2 2I i B1 and B2 are taken from table 3683.2-1
81 = B2 = 1 Hence = 6.57 + 11.67 = 18.24 kaimn
2
0 tot
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5. Allow'abl._tQe SA 376 Grads_ 316 S at room tuma. = 20 ksi - 14 kalmm2 Sm at 6501' (-3430C) - 16.6 ksi = 11.6 kg/mm Allowable stress *= 1.5 Sm (ASME III, article NB 36 52)
1.5 S = 21 ka/Irm 2 (room temperature)
M2
= 17.4 kg/mm (343°C)
6. Conclusion for orimarv stresses in the pine Since it annears that hot fluid has been carried by the pipe for a time of about 3 min, the hot allowable stress needs to be taken. Then it anpears that the actual stress is slichtly hicher than the allowable
18.24 > 17.4 ka/rrm 2 It should be noted that the fieiure of 18.24 k/zm 2 is a minimum, since it corresponds to the plastification of the support (M = 550.6 kar).
7. Primarv and Secondarv stresses in the mine The evaluation of secondary stresses (article NB 3653.1)
recuires the knowledge of the temperature gradients in the pine. It was thus not possible to evaluate these stresses.
8. Primarv stresses at the reducer Bending xontent Be = 1430s (385 - (405 - 13N) .; rm.
= 357 kcrm
1 t: - A-4 reducer 21£ " sch 1CC
OD = 2.875 in = 73.02 mm t = .375 in = 9.52 mm.
I3 3 I = 1.64-in = 26.9 cm
2 Pressure stress = rOD = 6.28 kd,/mm Bending stress = = 13.28 ka/rn2
2 Total stress = 19.56 kg/mm indicative since This stress should be considered more as force location.
it depends 6o much on the assumption of the The same conclusion holds as for the pipe stress.
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A
Direction of arobabl'.-
\efort.
Bolts (6 total) : Hexagonal head = 25 nun Damace : - no general distortion
- no rubbing evidence
- contact evidence in A
Figure 2 - Restraint R-1
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V4 VA w (mrk
/n o thc a 6 shos)
I'LOW -D KeeZ r, crt View A (marks on the shoes)
Damage : - top of insujation slightly smashed
- scratches on shoes as shown on view A
Figure 3 - Restraint R-2
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-Bolts 1 (4 total) m-10 *Damaae: - no evidenceat straps, pipe and
2 (4 total) ti-1 bolt- (I) ar.d (3)
3 ,4 total) pull out - all 4 bolts (2) .ave been strained
- gap measured as shown
.- rce = 41'/hclt - strain evidence in the r profile as 3Y
Fiaure 4 - Restraint R-3 II
- ri- P.. C4 L
QN X'.. Ut ,,1 .11. Q)
Figure 5 - Restraint R-5 Motion Evidence
.33
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1-dEZNAU - UNIT No 1 (NOK) v-'
STEAM DUMP FAILURE It4CIDL:;4T
Aug. 21, 74 PRESSURIZER RELIES LINE .
Figure 6 - Undamaged Relief Valve.
a Is B- : * : :: I .. 0 : ,)
STEAM DUMP F;_URE INC 'DE:T
Aug. 21, 74 PRESSURIZER RELIEF LINE
-. 5---- 7ej1o*
.5"; bj '- 4. -1.;
'1Ft Figure 7 Damaged relief valve General view showing the two fractured arms and the liefted operator.
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STEAM DUMP FAILURE LNCIDrE-JT
Aug. 21, 74 PRESSURIZER~RELIEF LINE
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Figur Daae Vle I De.ail of.fqaIsuu.,yoke
BEZIJAU - UNIT N* 1 . (NOK)
STEAM DUMP FAI1.URE INC I )ENT
Aug. 21, 74 PRESSURIZER RELIEF LINE
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\,LZNJAU - UNIT :-, I .~~\
STEAM4 DUMP FAILU1.,,. INCIDENT
Aug. 21,j 74 PPLESSURIZERJ RMTJLIe LZNE.
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. j j B1EZNAU - UNIT No I (I
STEAM DUMI' FAI J.' iIE INC TDL:'U'
Aug. 21, 74 PR^ZSSUR1ZE1I 1ULLIEF LINE
Figure 11- Support R4 (1)
General arrangenmcnt
100 x 50 x 5 profiles
50 Dun s lot t. on vi .
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I"Y
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STEAM DUMP FAILURE INCI Do'tNT
Aug. 21, 74 PRCSSURIZtR RELLEF LIIE
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STEAM DUMP FAlLURE INCIDENT
Aug. 21, 74 PRESSURIZER RELIEF LINE.
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.'. V.
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BEZNAU - UNIT No 1 (NON)
STEAM DUMP FAILURE.INCIDENT
Aug. 21, 74 PRESSURIZER RELIEF LINE.
I
Figure 15 - Ceiling Penetration (1)
(q-3
T
BEZNqAU - UNIT N* 1 (NOX)
STL1AMl DUMP FAILURE I:NCeIDENT
Aug. 21, .74 PRESSURIZER RELIEF LINE.
Figure 17- Ceiling Penetration (3)
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I I
NOX RFPORT ON P-uZNAU AccDr)ENtT 0F AUGTJST 20, 1974 TRIP TG-1/REACTOR TRIP/SI/
On Aucust 20, 1974 at 11:20 a.m. a trin on turbine TG-1 occurred 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 to operate, with the consequence that the relief valve did not open. That resulted in a rapid rise of coolant temperature, steam pressure and pressurizer level and pressure.
At 160 bar of pressure in the primary, the Pressurizer pressure relief valves opened, lowering raPidlv the Pressure in 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. Due to a disturbance, valve PCV-456, failed to close, resultinq in a lowering of RCS oressure up to 100 bar after about
1 minute. Reactor trinned resulting from a low pressure sianal (126.5 bar).
Due to the openina of the pressurizer relief valve, the pressure in RCS drooped to about 70 bar, corresponding to'
a saturation temperature of 284'C. Consecuentlv, steam appeared in the primary hot leg, filling the pressurizer.
Two or 3 minutes after trip, the operator recognised the failure of the relief valve and isolated it with the power operated valve 531. The water level began to dron, and
11 minutes after trip, automatic SI was initiated by low pressure ann level in the pressurizer.
14 . I !'
Pace 2. I.
SI systems worked normally and about 40 litres per second of water was soilled through the four SI pumo nozzles into the primary, causing a rise of pressure to 110 bars and a further rise of level to 70 %. The SI Pumas were then turned off and the Dower operated valves of the soray pipinqs were closed.
From that moment on, the pressurizer level could be controlle through charging pumps and release of steam, assumini the
. orimarv to cool down.
About 3 minutes after trio, the containment oressure alarm signal was actuated because of too high Pressure, and 1 minute later the high activity alarm. Maxim=m pressure in containment reached 100 mbar over normal. The operators activated the containment fan coolers. Since several safety alarms of the pressurizer relief t~ak were on, it was quickly assumed that the rupture disc was brokIen and that the 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 after TG-l trio. TG-2 was unreaular in behaviour, and the Position of the control valve retained constant during the pressure transient. The oerformances of TG-2 rose to about
214 MWe due to higher steam pressu-e (rise from 52 bar to 66 bar).
After TG-2 trio, following reactor trio, steam Pressure rose to over 70 bar, actuatina the safetv valves and thus lowerina Pressure to about 65 bar..
2. C(TROrOhLGICAL F
OST~N(~ S
August 20, 1974
Paae 3.
2.1. Reactor Trio Beginning of incident 11 h 20' 12"
TG-l main breaker off Pressurizer nressure low-trip 39,7n later Reactor trip breaker open 39,8' later TG-2 main breaker off 40,3" later SI actuation (pressurizer Pressure and level low) 11'55,9" later
2.2. Events as Recistered on Ai.arm Tvpoewriter
-
TIME
11:15 TG-1 power high 135,5 MWar
11:2C Allowable oil pressure of TG-1 too low
11:2C Pressurizer pressure 158.2 bar high.
11:2C Pressurizer pressure 159.9 bar high.
11:2] Tavq RCS-A hiqh 302.2*C
11:2]1 Steam nr. upstream of 66.3 bar TG-1 stop valve hiqh.
11: 2]L Tava RCS-A hich 305.20C
11:211 SG-A steam oressure 67.3 bar hich.
11:21L SG-R steam pressure 67.2 bar hiqh.
11:21L Steam or. upstream of 77.6 bar TG-l stoP valve.
11:211 SG-A steam pressure 73.3 bar hicih.
11:2:1 SG-A steam pressure 65.4 bar hiah.
11:2;a Safety oil nressure of TG-2 too low.
11: 2;2 Tavg RCS-A 285.2 eC
Paqe 4.
T:IMP
11:23 Steam pressure uostream of 68.1 bar TG-2 stop valve.
11:23 Pressurizer relief tank 62.86C
temperature hiQh.
11:24 Pressurizer level 79 '
11:24 Pressurizer level 88 %
11:24 Containment oressure hich 1.1 bar abs
11:24 Pressurizer relief tank level .20.2 %
low.
11:24 Pressurizer relief tank pressure 0.59 bar hiqh.
11:25 Pressurizer relief tank oressure 0.15 bar
11: 2; SG-A+3 steam oressures normal. 63.7 bar .
11:25 Containment activity high 17.3 mr/h
11:26 Loop B RCS flow low. 88 I
11:27 Containment air temoerature hiah 53.4 °C
11:32 Pressurizer level low. 6.8 '
11:32 Pressurizer level normal. 18 %
11:33 Surqe line temoerature too low. 271. 1C
11:34 Pressurizer levelthich. 58 %
2.3. Seauence of :Events for Pressurizer and Pressurizer Relief Tar.
TIME
11 h 20' 11.1" Pressurizer oressure above control ranae.
11.9" Pressurizer relief valve.
22.8" Pressurizer relief tank oressure hiah
23 .0" Pressurizer relief valve lcoked
23 .0" Pressurizer pressure normal
23.1" Pressurizer relies .ank le-ve' hiah 4.
24.2" Pres-urizer level hich.
33.o" Pressurizer relie-f tank oressure too hich.
35 .n Pressurizer Dressure under nornal.
.(ZO
. I
Paqe 5.
TIME
11 h 21' 00 . 4" Pressurizer oressure low - Trio.
01.2" Pressurizer pressure low - SIS
unlocked.
05.1" Pressurizer relief tank level hiqh.
13. 5" Pressurizer pressure low - SIS
unlocked.
11 h 233' 27.6" Pressurizer level hich - 1 channel t::
43 .3" Pressurizer relief tank level too him.
43.5" Containment pressure too hiqh.
47.1" Pressurizer relief tank level low.
11 h 24' 29.4" Pressurizer relief tank nfessure norma
51.2 Containment temperature hich.
11 h 25' 17.8" Containment activity hich.
3.' A%'qALYSIS OF O'FF CAUSES OF THE INCIDFNT
TG-l trioped due to hich casing vibrations, especially in casing 6. It had already been noticed that TG-l was sensitive to shocks. At the moment of incident, TG-l was set to function under maximum effort, so that it could support a maximum of vibrations.
The trio is not unfamiliar and would not have affected the primary 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 and air engine was broken, and probablv due to a dynamic effort on the pining at opening of the valve.
Consequently, the valve failed to close ar.d imitiated a raDid fall of pressure in nrimary. The pressurizer relief tank rupture disc broke, due to a mrolonced surce of orintarv coolant in the tank. Items 2 and 3 show the disc broke when the relief valve had already closed.
I I
Paqe 6 WATER COLLErCTED IN CONTAINMFE'r SUMP
Regen. hold up water Tank A 38 % - 100 9.8 m3 Regen. hold up water Tank B 16 % - 36 - 3.2 m
=
Total quanlity of water collected =13.0 m3 Pressurizer relief tank 80 % - 19 =11.2 m3
=
Water out of system. - 1.8 M
Since no further damage was noticed in containment, it could be assumed these 1.8 m3 of water were blown out.
4.1. Thermal Stresses in RCS
Beside a rapid water temperature rise of about 6C after TG-1 tripped, a rapid primary pressure rise fron 154 bar to 160 bar, there was also an imoortant temperature transient in area of SI nozzles. However, since the reactor's main pumps operated all the time, thus mixinq- cold spray water with hot coolant, it can be assumed that other components didn't underao high ternmerature gradients.
Furthermore, nozzle temperature and stress remained within design limits.
4.2. Damaaes to Relief Svstems During insp'action in containment after cooling of Drimarv, the following damaces in the pressurizer relief. systems were observed
- relief valve PLV 456 Mechanism broken on both sides and 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 7 It must be said that the relief tank is not desicned to accent steam from the Pressurizer for a Drolonqed time.
The damaqes to the relief valve is therefore a direct cause to the breaking of the rupture disc.
4.3. Turbines TG-1 The cause of vibrations to the casinc are most nrobably the stresses and shocks. The P sicnal from hydrogen seal oil svstem is due to casina vibrations.
Damaqes to the seal or casing are most improbable.
TG-2 a The oscillation from 172 MLWe to 110 MWe, and then to 215 M-We suggested that the bolts of the high pressure cylinder were loosened and had lost some of their tension.
A too small stress was noticed, due to leakaqe of the seals of the high pressure cylinder. Due to too hiqh rotational momentum at 215 MAe, the couplinq between turbine and generator was closely controlled.
5. When reviewing the sequence of events, the Failure of two systems, namely the steam duimb and the Pressurizer relief system, we came to the conclusion that it did not brinc to an uncontrolable nor a damaqinq situation. nurina the incident, no activity (in gas or liquid form) in the surrounding area reached an uncontrollable level.
The generator safety valves maintained the steam pressure within allowable limits. The SIS broucht back the Primarv to a safer pressure, allowinc normal cooldown conditions.
6.' PROPOSAL FOR MODIFICATTONS
6.1 Control of cenerator 1 Generator 1 reaching ranidclv to casinq vibrations,it will
Paqe 8 be tried to see if the regulator can be modified in order to have a quick action.
6.2. Pressure Reaulator Tests will be made to see if the first row of impellers in the pressure regulator of the turbine must not be reviewed in order to limit power to 190 MWe.
6.3. Steam DumD System a) Revisions and calibrations should be made in Ateam duff system (before opening of steam dumo valve.)
b) Studies will be made, to make periodic controls of steam dumo while in operation. It should helo to.insu--.
better safety limits (for example : unwanted oneninq c.1 steam dump valve).
c) A control type writer linked to the steam dumo will be installed in order to control the opening of steam dumo valves and to check the qood working of oil OUmos .
6.4. Pressurizer Relief Svstem The first measure to be taken, is to reoair the damaced valve, 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 secured and how shock at opening of relief valve can be avoided are further measures to he taken.
W-
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHWINGTON, D. C. 205
-YIS1979 MEMORANDUM FOR: D. F. Ross,.Jr., Deputy Director, DPM
FROM: Ashok Thadani, Task Manager SUBJECT: STUCK OPEN POWER OPERATED RELIEF VALVE AT FOREIGN PWR
In the process of gathering data on power operated relief valves (PORVs)
for our report on Westinghouse plants, we were informed by Westinghouse that they were aware of only one instance of a PORY failing to reclose after opening. No failure of this nature had been observed on any U.S.
reactor plant designed by W. The failure, according to W, occurred at one 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 not Seen observed on any U.S. Westinghouse reactor.
To follow up on the apparent foreign reactor PORV failure, we contacted Howard Faulkner of NRC International Programs and informed him of our need for additional information. Our basic need was to determine whether this 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 Scott Newberry) and the Swiss Federal Office of Energy of Switzerland was ar- ranged for the morning of May 15 to obtain this information. Howard Faulkner informed the Swiss that we would treat this information as con- fidential and would telecopy them a copy of what we intended to include in our W evaluation report prior to its issuance.
A sequence of events for the turbine trip and associated PORY failure to close described by Mr. F. Weehuizen, Head of Energy Section, is attached.
We requested additional information to supplement that in the phone conversation:
1. Event reports pertaining to the event
2. PORV description, manufacturer and failure mode
71 i/2=oo73
I
U. F. Ross, Jr. 22-- 1979 Based upon this phone conversation, we note that:
1. As demonstrated by this event, pressurizer level will remain above the trip set point for ECCS actuation for a stuck open PORV.
ECCS did not actuate automatically until the operator shut the PORY
isolation valve.
In this case we do not know how soon the coincident signal (Lo Level/
Lo Press.) would have automatically initiated HPI and the subsequent operator actions since the PORY was isolated atiminutes.
2. The indications in the control room of actual PORY position and relief tank parameters appear to have provided the operator with sufficient information to make a reasonably rapid assessment of the problem and take appropriate action.
Since this event occurred about five years ago and because of its relevance on our current deliberations on W designed plants, we recommend that complete information package including plant data be obtained and reviewed, as well as the role of the operator.
We therefore recommend that all operating Westinghouse reactors modify the pressurizer level/pressure coincidence ECCS actuation as already directed by I&E bulletins 79-06 and 79-06A and that we continue to pursue the 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 more information on turbine bypass system failure modes as a lower priority consideration.
A. Thadani Task Manager cc: E.G. Case R. Mattson L&L. Tedesco XT. Novak lf7.Faulkner S. Newberry
Enclosure
1. 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 a controller malfunction caused by operator error during previous maintenance period.
3. Primary system temperature, pressure and pressurizer level increase. PORV opens.
4. Primary pressure decreases. After 10 seconds PORV should have shut but remained open.
5. Reactor trip on low pressure (pressurizer level still above low level trip, therefore ECCS has not yet actuated on coincident low pressure -
low level)
6. Reactor Coolant System pressure decreases to saturation. Voiding in hot legs. Operator observes flow oscillations and reactor coolant pump vibrations. He did not trip the reactor coolant pumps.
7. 2-3 minutes after the reactor trip, the PORY isolation valve was shut by the operator. He had received increasing pressure and temperature indication in pressure relief tank. He also had open indication of PORV (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, ECCS
actuated on coincident low pressure/low level ECCS performed as designed
10. Pressure increased to 110 bars ('1600 psi).
Pressurizer level increased to 70% of indicated range.
Operator tripped HPI and maintained pressurizer level using charging 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,<
UNITED STATES
in NUCLEAR REGULATORY COMMISSION
WASHINGTON, D. C. 20555 SEP 13 1979 MEMORANDUM FORi H. R. Denton, Director, NRR
E. G. Case, Deputy Director, NRR
D. Ross, Deputy Director, DPM
R. Mlattson, Director, DSS
FROM: Darrell G. Eisenhut, Acting Director Division of Operating Reactors SUBJECT: INCIDENT AT BELGIUM DOEL 2 REACTOR
In response to our following up on a rather large, sudden steam generator tube rupture at the Doel 2 nuclear power plant in Belgium, we have received the attached report. You may find this incident particularly interesting since the unit underwent a transient where pressurizer level apparently went offscale high. Strip chart recordings of the event are enclosed.
We hope to be obtaining more informa ion on this event in the near future 0
Darrell G. Eisenhut, 4cting Director Division of Operating 'Reactors Enclosures:
As Stated cc: S. Hanauer F. Schroeder B. Grimes P. Check G. Lainas S. Levine V. Stello W. Russell
_.Q CENTS0 DETUDE DE L'ENERG/ NUCLEAIRE
- ,C.E.N. I S.C.K.
Jim .. *ug..'Yd tW 0;ar .uW6euS
r VeuiI'ei adressr votre fipones Mr. Joseph D. LAFLEUR, Jr.
on doux examplsres aua Deputy Director LABORATO0IES DU C.E.N /S.C.K. Office of International Programs Bouretang 200 B. 2400 MOL UNITED STATES NUCLEAR REGULATORY COMMISSION
le.(0p 01e X2 1e WASHINGTON D.C. 20555 elex SCKCEN-Moi 31922 SEP 19793L U S AC
Adr. twegr.: Centratom Mol I. U.S.A
MOL. is 21.08.79.
V Ilnre V/rnfI. Nrif.
Centrale BR3 FM./mb
5.5126/71 Dear Dr. LAFLEUR,
As a first answer to the telex of Mr. H.J. FAULKNER
NRC-BHDA, dated 8.8.79, I send you here enclosed a report describing the steam generator leak incident at the Unit 2 of the Doel nuclear power plant.
This report has been transmitted to me by "Tractionel Engineering", a division of the compagny "Societl de Traction et d'Electricite" in Brussels ; as you most probably know, this division is playing the role of engineering office for the benefit of the Doel plant operator compagny (EBES).
I hope you will find in this report satisfactory answers to all your questions ; do not hesistate to ask for eventual additional informations.
Yours sincere .
F. MOTTE
BR3 Plant Superintendent.
Enclosure : "Report on the incident at Doel 2 nuclear power plant Severe leakage in steam generator B on June 25, 1979".
t10
PD/vEF
20.07.79 SEP 3 1979 REPORT ON THE INCIDENT AT DOEL 2 NUCLEAR POWER PLANT
SEVERE LEAKAGE IN STEAM GENERATOR B ON JUNE 25, 1979.
1. STATUS OF THE POWER PLANT AT THE MOMENT OF THE INCIDENT
The primary system was being heated up after repair works at the actuation system of the main steam valve.
At the moment of the incident, temperature in the primary system was - 2551C (refer to point A on Fig. 1 & 2) and pressure had reached its rated value of 157 kg/cm2 (refer to point A on Fig. 3 & 4).
Tne reactor was subcritical with all rods in.
Secondary pressure in the steam generators was - 45 kg/cm2 the saturation pressure corresponding to 255'C (refer to point A on Fig. 6 & 7).
For some time, A-loop steam generator had shown a low activity value along the secondary side (below admissible limits) that indicated a small leakage.
2. SEQUENCE OF THE EVENTS (refer also to various computer data given in attachment)
2.1. Initiating phase About 7:20 PM, a quick pressure decrease is recorded in
2 the 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 confirmed that the relief valves are closed and their isolation valves are preventively closed. The level in the pressurizer quickly decreases (see Fig. 5) and the electrical heaters are automatically disconnected.
At the same time, a quick level increase is recorded.in B-loop steam generator (see Fig. 7 point B). The activity measurement channels of the blowdown system record a maximum value.
The combination of all those signals indicates a severe leakage in B-loop steam generator. The faulted steam generator is then immediately completely isolated along the steam side and the discharge valve to the atmosphere is set at maximum pressure.
Meanwhile -the third charging pump is started (was set apart to be maintained) , but the three charging pumps are not sufficient -to compensate the loss of fluid in the steam generator. Indeed, the CV tank is readily empty and the charging pumps are automatically supplied from the 2R11 refuelling water storage tank. To increase the subcooling primary pump B is stopped and letdown starts througb A-loop steam generator (see Fig. 3, point B).
2.2. Actuation of safety injection About 20' after the incident started, the threshold pressure
2
(118.5 kg/cm ) to actuate the safety injection is reached.
The emergency diesels start within the required time lapse but are not necessary. Phase A isolation and ventilation isolation of the reactor building are achieved. The vital components not yet in operation are started.
j .
When reaching the 108 kg/cm value, all HP SI-pumps discharge into the primary system, and the pressure decrease is stopped (see Fig. 3, point C).
To prevent the secondary pressure in the faulted steam generator from reaching the opening pressure of the safety valves, the primary pressure is successfully decreased (see Fig. 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 quickly increases and it fills up completely (see Fig. 5). Spray- is temporary stopped and pressure stabilizes at zero flow pressure of HP SI-pumps.
The automatically started auxiliary feedwater supply results in a pressure decrease in B-loop steam generator (see Fig. 7, point C). The auxiliary feedwater supply pump of the faulted disconnected steam generator is locally stopped and isolated (Fig. 7, point D). This cannot be performed from the control room since the SI-signal stillprevails. The auxiliary feedwater supply tank is filled up from Doel 1.
2.3. Cancelling of SI-signal Pressure decrease was now mandatory a) to avoid the opening of safety valves of the faulted steam generator.
b) to start, as soon as possible, the shutdown cooling system (low pressure circuit 1) to stop the letdown of slightly contaminated steam through the A-loop steam generator.
4.
Firstthe safety injection signal had to be cancelled.
This had to be performed more than once (each time requiring
5 minutes interval) because of a relay fault.
After definitively cancelling the SI-signal, two HP S%-pumps are stopped and soon thereafter a third one (Fig. 3, point F).
While considering the subcooling margin, the last HP SI-pump is stopped. Pressure successively decreases to reach -
65 kg/cm 2 (Fig. 3, point H) (saturation pressure is 4 15 kg/cm 2 at that moment).
It is then tried to initiate the CV-discharge line, but valves do not open. Some time goes by before t-he reason therefore is determined. Due to phase A isolation there is no longer a-compressed-air supply in the reactor building. After re-opening the compressed-air supply line the discharge'line is opened (Fig.-3, point I). Pressure
_ ...
decreases, first quickly, then slower. ..'.
The loss of compressed-air supply has also resulted in the closure of CC-valves to the primary pumps. The pumps have run for a long-time without cooling of the thermal shield, however without alarm temperatures were reached.
2.4. Initiation of the residual heat removal system As the CV-system permittted only a slow pressure decrease, X15-
S.
the interlock, which maintains the isolation of the RHRS
kg/cm2 up to a pressure of 28 kg/cm , has been bypassed at 31 There was indeed a sufficient margin compared to the design
2 pressure of thesystem (42 kg/cm ). Thanks to this operation the letdown through A-loop steam generator could be stopped earlier and the discharge of slightly contaminated steam could be reduced (Fig. 3, Point J).
2.5. Further sequences The abovementioned operation allowed a primary pressure decrease below the value of secondary pressure in the faulted which B-loop steam generator. The secondary level decreases, creates a dilution risk. The boric acid concentration is controlled every half hour (stabilized howerver at + 1500 ppm).
Thanks to the cooling down, pressure decreases slowly in the B-loop steam generator and reaches a value lower than primary pressure. From this moment on, attention is paid in to always maintain the primary pressure higher than that the steam generator.
Despite the cold water so discharged in the steam generator, of pressure goes on decreasing slowly (due to the presence a warm water film at the water surface).
the As the level of water in the steam generator approaches upper limit of the broad level measurement pressure is sufficiently low (+ 12 kg/cm2) to inject nitrogen.
liquid The secondary drain line is coupled with system B for waste, and the steam generator discharges into It through the nitrogen pressure.
The nitrogen is only slightly contaminated after this and and can be discharged via the annulus between primary secondary containments.
I1 i
A, '=6.
.2.6. Comments and conclusion The incident has been handled as proscribed and no damages have 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 air supply in the reactor building.
Attachment 1 - Computer data
1. Initiating phase
19 21'06" pressurizer pressure below reference pressure
- 9 22'51" demand for charging pump higher speed
19 23'31" disconnecting pressurizer heaters by low level
19 23'32" CV letdown station valves closed valves and
19 25'42" closing of isolation valves of relief spray valves
19 26'14" low pressure in primary system
19 30'30" very low pressure in pressurizer
10 30130" high level in B steam generator
19 38'32" B primary pump disconnected
2. Safety injection phase
19 40'18" low pressure in pressurizer
19 40 '19" safety injection through low pressure in pressurizer
19 40'19" diesels started
19 40'19" reactor building ventilation isolation
19 40'20" phase A reactor building isolation
19 40'24" actuation signal HP SI.-pumps
19 40'33" HP SI-valves opened
19 43'28" very large auxiliary feedwater flow to A SG
to B SG
19 44'39" very large auxiliary feedwater flow
19 53'12 auxiliary feedwater supply pump B disconnected supply tank
19 56'37" very low level in auxiliary feedwater
19 57'11" pressurizer level normal
19 57'29" pressurizer heaters re-started
19 58'48" high level in pressurizer
3. SI-signal cancelling phase cancelled and
20 00'15" automatic starting signal of diesels SI-pumps starting signal cancelled
20 00'21" back to SI
2.
20 03'24" LP compressed air in reactor building
20 05'59" safety injection ordered
20 06'05" safety injection
20 10'59" reactoF building ventilation isolation ordered
20 21'15" HP SI-pump B disconnected
20 25'22" HP SI-pump A disconnected
20 38'33" valve CC 096 closed
20 40'25" valve CC 099 closed
20 48'54" compressed air supply to reactor building restored
20 49'00" primary pumps CC-valves re-opened
4. Actuation of RHRS
22 35.54" valve RC 003 opened
180
- i A j 'WAA%
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S4 4 &
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2. Ro. P. S 4A - 6A DruX SGA 0 - 85 kg/cm,2.
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1. Bi. AL. FnW 9 B - 10 B Peil SG B 0 - 3500 am cc., .2Pjt 6i D4dc
2. Ro. P. HES 4 B - 6 B DruX Sd Z 0 - 85 kg/cr2 meting in dienst spoor I :
. spoor 2 :
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' 1 7, 'WAEO71 (CO40)(1-096C92442)fPO 08/30/79 0640
ICS IPmIHA I l.3SS
IISS Ff, 'Ul 30 O640
P?;*S O.iY COV'.iISSIOi\ '. ASHiNGTON1 OC
U'! 9.1! 3 FI3 ?2 1 TG3 /229 lU'J.'!W- CO BEAN 13?
8.NTW^RP~ TELFX 132/125 30 1017 PI/S.R.
iR JOSEPH D LAFLEURJ JR
)FPLJTY D)IRECTOR
OFr1CF Or I\TFRNATlON:AL PROC-RAMS
U 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 PLEASE
T; F ¢FT-i SPE.CIFIC Ai'St'!E..S TO ThE FOUR OJE3TlONS kRAISL RY
YOU.R ). F4ILcNF:R ON THrlF .JOFL 2 STEANM GENFic~ATOR I.NCIDEtN'T
1s TiRE MAC-N1TULDE
OL (Q0555) 5.516/71 21 2 1.
3/229 -:R JOa'?. 0 LAiL-. .J.7 JN Pe/59 oT i--- LFAk ::.As ESTIMAT-D AS .49OUT 30 TONS/eOU.. ANr
-)V: L'j? RAOPIflLY
S. TF-E LEak IS LOC.RTEL, ON THE TOP OF PIPE NR 1/?hs OF STEA^. GENER.ATOR
3 007L ° IN Thr -TRA--OS OF Th' U-$ENli)
3. 3UoP.CTED CA'JZ' STRFSS-CORROS10N DU;:. TO OV4LIZATION
h. !rENTINP MAXIMyUM 450 MICR9NETFA.R
NCt rLOW
COL 3) S. 1/?4 3o. A. 450
.31/?9 il. JOSFPI-. D L47L7UR JR P3/25 SLOT DEFOR ATION AT ALL NO TU?EF i:ALL THMNINJ FOUNfl THESE ANSVERS, WERE FO.Ri;ULATEFi .3Y ThE DOEL I .AiND °
C. -
PLANT SUPERINTENDENT -'C
YOURS SINCEP*ELY
F MOTTE
COL 1 q
.RETPl Mr'St:
NN
v,.U Tl'.X '.!SAH
0
Mr. William J. Cahill, Jr. 50_3 Consolidated Edison Company of New York,_Jnc. 50-247 cc: White Plains Public Library
100 Martine Avenue White Plains, New York 10601 Joseph 0. Block, Esquire Executive Vice President Administrative Consolidated Edison Company of New York, Inc.
4 Irving Place New York, New York 10003 Richard Remshaw Nuclear Licensing Engineer Consolidated Edison Company of New York, Inc.
4 Irving Place New York, New York 10003 Anthony Z. Roisman Natural Resources Defense Council
917 15th Street, N.W.
Washington, D. C. 20005 Dr. Lawrence R. Quarles Apartment 51 Kendal at Longwood Kennett Square, Pennsylvania 19348 Theodore A. Rebelowski U. S. Nuclear Regulatory Commission P. 0. Box 38 Buchanan, New York 10511 John D. O'Toole Assistant Vice President Consolidated Edison Company of New York, Inc.