ML20058B378
| ML20058B378 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Browns Ferry, Salem, Oconee, Sequoyah, Arkansas Nuclear, North Anna, Ginna, San Onofre, Maine Yankee, Rancho Seco, Zion, Fort Calhoun, 05000000 |
| Issue date: | 07/06/1982 |
| From: | Michelson C NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | Lafleur J NRC OFFICE OF INTERNATIONAL PROGRAMS (OIP) |
| References | |
| NUDOCS 8207230690 | |
| Download: ML20058B378 (54) | |
Text
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JUL 6 1982 i
f2N MEMORANDUM FOR:
Joseph D. Lafleur Jr., Deputy Director Office of International Programs FROM:
Carlyle Michelson, Director Office for Analysis and Evaluation of Operational Data
SUBJECT:
IRS REPORTS Please forward the following enclosed IRS reports to Mr. Otsuka of the NEA.
1 Bolt Corrosion 2.
Loss of 125 V DC Bus 3.
Inadvertent Containment Spray Actuation 4.
Contaminated Air Systems 5
Investigation of the Relative Frequency of Valve Overtravel Anomalies 6.
Loss of Shutdown Cooling and Positive Reactivity Addition 7.
Failure of Main Steam Isolation Valve to Close Due To Failure of Two DC Solenoid Valves to Actuate 8
Load Reduction Transient 9.
Failure of the Main Transformer Original Signed by Carlyle Michelson Carlyle Michelson, Director Office for Analysis and Evaluation of Operational Data
Enclosure:
As stated Distribution CentralFile/:'
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OECD NUCLEAR ENERGY A
o AGENCY 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telex 630668 AEN/NEA RESTRICTED DIFFUSION RESTREINT
.N o J R_S__.
Title - Titre Failure of Main Transformer Country - Pays Date of Incident - Date de l' incident United States
.lulv 25. 1981 Type of Reactor - Type de reacteur PWR Plant - Centrale Licensee - Dstenteur du permis d'exploitatic North Anna Viroinia Electric and Power Company
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Unit N*
- Tranche n*
g Westinghouse Power - Puissance First Commercial Operation -
3g,99
""'I"'*I
- ' "I
'" ''I
890 Systems or Components Affected - Syst mes ou composants affectss Main Transformer Initial Plant Condition - Etat initial de la tranche 0% power, Shutdown Way in which Incident was Detected ?
Comment l'inciderrt a-t-il sts dstects ?
Transformer Failure Rad: tion Exposure or Radioactivity Release -
Exposition aux rayonnements ou libsration de radioactivits Date of Receipt - Date de rsception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, :r.esures prises ou projetees et
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f Event Description On July 25, 1981, the phase C main transformer failed two minutes after it was energized from the 500 kV system. The licensee concluded that the initiating cause of this failure was an incipient failure in the transformer coil resulting from failure of the phase B transformer on July 3.
(Describe.d in a previous IRS report, " Fire Resulting from Transformer F~ailure.")
In
~ addition, the July 25-failure was the fourth such failure since November 1980.
All four of these failed transformers had been in service on the Unit 2 main generator output distribution system at the time of their failure.
The C phase main transformer experienced an apparent high-voltage winding-to-ground failure which tripped the switchyard 500 kV breaker, thus, isolating the damaged transformer from the balance of the station 500 kV distribution.
The fault caused extensive damage to the transformer rupturing the casing in several places and damaging the high voltage output bushing and adjacent lightning arrestor insulator. The oil contained in the transformer was spilled onto the gravel surrounding the transformer base. There was no fire during this event.
This transformer was a replacement obtained from another utility because a previously failed transfornier had been returned to the vendor for repairs. The handling and preoperational service procedures performed on the f ailed C main transformer 'did not disclose any major discrepancies.
Cause of Event A task force comprised of the licensee and Westinghouse (transformer manufacturer) personnel was established to study the. transformer failures. The four failures were characterized by two failure modes.
Failures 1 and 4 were winding to ground faults.
Failures 2 and 3 were high voltage bushing to ground failures. All of the failures apart from failure 2 involved high voltage to low voltage windings.
The failed Unit 2 transformers shared a rather unique background:
l 1.
They had all been handled and shipped several times before being placed in service in Unit 2.
2.
They had been in a bank of transformers which had experienced a failure of at least one transformer at least once prior to their own failure.
3.
The high voltage buthing for these transformers had also been handled more than usual and may have been stored improperly.
Improper storage of 'the high voltage bushing coupled with an over-voltage condition could cause failure of the bushing as experienced in failures 2 and 3.
4.
The transformers associated with failures 1, 2, and 3 had been subjected to several documented over-voltage transients.
The transformer associated with failure 4 had been subjected to an over-voltage condition of unknown magnitude and short duration on the low voltage side during failure 3.
t
. It was concluded that failures 2 and 3 were a result of the high voltage bushing failing due to a combination of improper storage and over-voltage.
Failure 1 resulted from a combination of insufficient cooling, mechanical relief device operation, gas pressure control failure, and inoperative transformer alarms, which allowed nitrogen gas to exist in the transformer 011~. ' Thi's~ gas-in-oiT combinatioh ' reduced the dielectric strength enough to cause failure at operating voltages. The fourth failure actually began during the time the third failure occurred. The initiating fault did not lead to the failure of the fourth transformer sinultaneously with the third because the protective relaying de-energized the transformer bank before the fourth transformer fault had grown sufficiently. When the fourth transformer was subsequently back fed, the previously initiated fault in the low voltage winding caused a catastrophic failure to occur after about two minutes.
Reason for Reporting to IRS This event is reported under category 3, "Significant Deficiencies in Design, Construction, Operation or Safety Evaluation" which could lead to a loss of a required safety function.
Actions Taken The existing transformers in the Unit 2 output bank, their high voltage bushings, insulating and cooling oil purity, associated on-site high and low voltage distribution components, protective relaying, alarms, and instrumentation were thoroughly tested and accepted for operation.
In addition, those site procedural and material matters which could have contributed to the failures (e.g., automatic operation of the transformer oil coolers, winding temperature indication and transformer alarms operable) have been corrected, l
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OECD NUCLEAR ENERGY T
'g D
AGENCY
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. 38, bd. Suchet 75016 Paris INCIDENT REPSRTING SYSTEM Tel. 524.96.93 Tele:: 630668 AEN/NEA RESTRICTED Na. IRS DIFFUSION RESTR.EINTE Ti t-l e - Titre-Load Reduction Transient 9
Country - Pays Date of Incident - Da t'e de l' incident United States Ja nuary 1,4, 1982 Type of Reactor - Type de tsacteur PWR Plant - Centrale Licensee - Ddtenteur du permis d' exploitation Salem Public Service Electric & Gas Co.
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Unit N*
- Tranche n
- 2 Westinghouse Power - Puissance
.First Co*mmercial Operation -
1115
""*l"*#l d' "I"*
- "I'*
June,1977
- Systems or Components Affected - Systbmes ou composants affectds Reactor Coolant System, Steam Dump System, Turbine Initial Plant Condition - Etat initial de la t,r a n ch e 94%
Way in which Incident was Detected ?
Comment 1 *-inciden t a-t-il Etd ddtectd ?
Steam Generator Feed Pump Alarm.
Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou libdration de radioactivitd None Date of Receipt - Date de rsception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetses et j
enseignements tirss (signification de l' incident pour la sGrets) doivent figurer j
sur les pages suivantes.
o Event Description Oa January 14, 1-982, a turbine load reduction was initiated at Salem Uni t 2.
This was in response to a steam generator feed pump low suction pressure which was apparently caused by. a secondary system disturbance associated with the No.- 2A feedwater heater and moisture separator reheater drain tank level control system.
In conjunction with the manually initiated load reduction, the operator also bypassed the condensate polishing system.
In order to reduce the RCS Tave, the operator tried to insert the control rods in Bank D but failed to do so because of a failure of the firing circuit control card in the power cabinet. He then manually initiated RCS boration.
As a result of the turbine load reduction, the condenser steam dump system had been armed in the load rejection mode of operation. There was a large mismatch between the reactor power (92%) and the turbine load (50%) and this power mismatch was being rejected to the condenser by the steam dump'ted system. The operator increased the turbine load and the dump valves star to modulate closed and Tave was being held stable at 5800F.
The operator reset the steam dump system load rejection signal causing the steam dump valves tc close. This resulted in Tave to peak. at 5920F _
causing pressurizer level to increase from 54% to 78% and pressure from 2200 psig to 2325 psig. The pressurizer pressure was reduced by the operation of both pressurizer spray valves. RCS temperature was decreased-400F in about 15 minutes.
The rapid increase in RCS Tave was also reflected back into the secondary system causing the steam generator pressure to rapidly increase resulting
-- in-lifting one of the steam generator safety. valves. The safety. valve _
failed to reseat fully due to the lifting disc associated with the manual lif ting arm jaming the valve travel. The safety valve was still blowing out but had no effect on unit stability. The safety valve was reseated four hours af ter it had opened by removing the janmed lif ting arm. The valve.was repaired and the unit was maintained at 46% steady power. _ _,.
Cause of Event The event was caused by a reduction in the steam generator feedwater pump suction pressure. This was further complicated by the inability to insert the control rods due to the failure of a firing circuit control card in the power cabinet.
Reason for Reporting to IRS This event is being reported pursuant to criteria 2.4, " Degradation of Systems Required to Control Criticality."
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_2 Actions Taken Station procedures are being revised to provide more guidance should a similar incident occur.
1.
In the future, Tave will be confined between 5410F and 5810F.
Outside these bounds, the unit will be tripped.
2.
For a large power load mismatch, the unit will be tripped.
3.
If a cooldown limit is exceeded, safety injection will be initiated.
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OECD NUCLEAR ENERGY T
g D
AGENCY
{
_j 1,s 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Tel:x 630668 AEN/NEA RESTRICTED DIFFUSION RESTREINTE e.No.lRS Title - Titre Failure of Main Steam Isolation Valve to Close Due to Failure of Two DC Solenoid Valves to Actuate Country - Pays Date of Incident - Date de l' incident January 15, 1981 United States Type de tsacteur Type of Reactor PWR Plant - Centrale Licensee - Dstenteur du permis d ' exploitation Zi on Commonwealth Edison Company
' ' ~
Unit N*
- Tranche n*
1 Westinghouse Power - Puissance First Commercial Operation nwe(net)
D te de mise en service October 1973 1040 Systems or Components Affected - Systames ou composants affect 6s Secondary Cooling Systen, DC Solenoid Valves, and Main Steam Isolation Valve Initial Plant Condition - Etat initial de la tranche Hot Shutdown Mode and Borated to Cold Shutdown Conditions l
l l
Way in which Incident was Detected ?
Comment l' incident a-t-it sts ddtects ?
l Operator observation 1
Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou lib 6 ration de radioactivits l
l None l
Date of Receipt - Date de rsception Date of Distribution - Date de distribution description, possible causes, actions taken or planned and lessons learned Event i
(safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetees et m.
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Event Description On January 15, 1981 while in the hot shutdown mode and borated to the cold shutdown conditions, the refueling closure time testing of Unit 1 main steam
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isolation valves (MSIVs) was performed.
1B MSIV failed to close upon manual actuation from the control room..0perability of the MSIVs is a limiting I
condition for operation. The valve was closed within an hour.
~Cause Loop B MSIV failed to close due to concurrent failures of two DC solenoid valves (Teledyne-Republic Model 21110-6202-5200'and Keane Control Model 132 SS-110-02C-EM) to actuate. There are two independent trains of valves that can cause closure of the MSIVs.
Each DC solenoid valve is on a separate train. The function of the DC solenoid valve is to allow fluid to f' low to the pilot operated check valves of their train, which opens the check valve and causes the MSIV to close.
The Teledyne solenoid energized upon actuation from the control room, but the valve failed to shift.
This failure of the valve to shift is due to the impurities in the oil settling out on the valve surface over the months. the valve lays idle. This has been a recurring problem with Teledyne valves due to the large surface area of their movable spool piece that is wetted by the oil.
The Keane valve solenoid fa' 1 energize upon actuation from the control room
.lesults'from the Keanne investigation indicated and failed to actuate manu
- the solenoid coil shorted c-_1ng the valve to overheat and seize.
Reason for Reporting Thi's occurrence is reportable to the IRS pursuant to criterion I.4,.'ILoss of Containment Function or Integrity," criterion 3, "Significant Deficiencies in Design,- Construction, Operation, and Safety Evaluation," and criterion 4, "Significant Generic Problems."
Actions Taken The last four Teledy,1e valves were replaced with Keane valves this outage to finish a modification that resulted from Teledyne problems. The failed Keane valve _has been replaced with another Keane valve and the operability of the MSIV Since this was the first failure of a Keane valve, no further has been proven.
action is deemed necessar.y at this time.
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{y OECD NUCLEAR ENERGY, A
D AGENCY
.~
3e, bd. 5uchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telax 630668 AEN/NEA RESTRICTED
- NS.lR$
DIFFUSION RESTREINTE Title Titre i
1 Loss of Shutdown Cooling and Positive Reactivity Addition.
o Country - Pays Date of Incident - Date de l 'inciden t March 15, 1982 United States Type of Reactor - Type de r6acteur PWR Plant - Centrale Licensee - D4tenteur du permis d' exploitation San Onofre Southern California Edison & San Diego Gas & Electric Manufacturer - Fabricant Unit N*
Tranche n
- 2 Combustion Engineering Power - Puissance First Commercial Operation 1100 MWe(net)
Date de mise en service NOt Ett COI3MBTCb51 Systems or components Affected - Syst mes ou'composants affectss Reactor C,oolant System and Refueling Water Storage Tank Initial Plant condition - Etat initial de la tranche Core fully loaded with unirradiated fuel Way in which Incident was, Detected ?
Comment l' incident a-t-il ses dstects ?
Plant operators noticed lack of flow in shutdown cooling train in use I
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Radiation Exposure or Radioactivity Release -
l Exposition aux rayonnements ou libsration de radioactivits None Date of Receipt - Date de rdception Date of Distribution - Date de distribution l
Event description, possible causes, actious taken or planned and lessons learned (safety significance of incident) should be included in the following pages, h memRcus reRAe@e ou grojetees et
Event Description With the reactor core fully loaded with unirradiated fuel, plant operators noted that there was no flow in the shutdown cooling (SDC) train then in operation.
The redundant train was placed in operation but no flow was obtained from the train.'
The operators opened the
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pump su'ction valvbs to the Ref0eling Water Storage Tank (RWST) to reestablish pump prime.
This operation was successful and the required flow was reestablished.
The boron concentration in the RWST was less than that in the Reactor Coolant System (RCS) and, as a result, the concentration in the RCS was reduced, although it was still well above the Technical Specifi-cation (TS) limit.
This dilution, however, corresponded to a reac-tivity addition of about.64 percent which exceeded the TS reporting limit for the subcritical condition (0.5 percent).
Cause of Event The boron concentration in one of the two Refueling Water Storage Tanks varied from about 612 ppm at the top of the tank to about 1900 ppm at the bottom of the tank where the boron concentration is usually measured.
Prior to this event this tank had been filled with borated water from a portable batch system.
When this water was found to be contaminated with iron, the tank was emptied and flushed out with clean water.
It was then filled about 15% full of clean water before borated water was
'added.
This tank admits' recirculation water in the bottom of the ta.nk and water is drawn from the bottom of the tank.
This inadequate mixing i
design, coupled with the dilution of the boron concentration in the RWST
^
by the leaking of approximately 17,000 gallons of water into the RWST, resulted in both a relatively low baron concentration and in a stratifi-cation of boron concentration.
l It is believed that the loss was due to gas binding of the SDC pumps while back-flushing the letdown system filters with high pressure l
nitrogen with the volume / control tank bypassed so that high pressure nitrogen could enter. into the.SDC piping.
This hypothesis requires... _
several operator errors, some of which are due to poor human engineering features.
Reason for Reporting This event is reported under criteria 2.4 " Degradation of Systems Required to Control Criticality"and criteria 3 because it indicates a major condition not specifically considered.
Actions Taken No specific action was taken because the configuration used while backflushing the letdown filter is unique.
Under normal circumstances j
the event would not have occurred.
OECD NUCLEAR ENERGY T
D AGENCY
{
.,s 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Te1. 524.96.93 Telex 630668 AEN/NEA RESTRICTED No.lRS DIFFUSION RESTREINTE Title Titre Failed Air Supply Solenoid 0-Ring Disables North Salt Water Cooling Pump Country - Pays Date of Incident - Date de l' incident United States March 10. 1980 Type of Reactor Type de tsacteur PWR Centrale Licensee - Dstenteur du permis d' exploitation Plant San Onofre Southern _ California Edison and San Dieao Gas and Unit N*
- Tranche n *
" " '# ' ~ ; ~ ' ~ I"
- A##"
1 Weetinnhnnen Power - Puissance First Commercial Operation -
436 M We (n e t )
Date de mise en service danuary 196S Systhmes ou composants affectis Systems or Components Affected Salt Water Cooling System Initial Plant Condition - Etat initial de la tranche 100% power Way in which Incident was Detected ?
Comment l' incident a-t-iL 4td dstects ?
Lack of flow indication on meters Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou libdration de radioactivits Date de rsception Date of Receipt Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, cesures prises ou projetnes et
Event Description On March 10, 1980, while operating at 100% power, this unit experienced a complete. loss of the salt water cooling (SWC) systen. The salt water cooling system is the ultimate heat sink for the component cooling water which cools ciertain safety-related equipment.
C;iuse of the Event The event involved an unlikely triple failure wh'ich resulted in exceeding the plant's limiting conditions for operations and was later determined to be an abnormal occurrence. The equipment failures were:
(1) shearing of the south salt water cooling pump shaft; (2) failure of the north salt water cooling pump discharge valve to open; and (3) failure of the auxiliary salt water cooling punp air priming system. The second failure was due to a failed air supply solenoid 0-ring. The solenoid 0-ring is believed to have failed due to the abrasive action of desiccant which had migrated through the instrument air system to the valve.
In addition to finding desiccant in the air system, the licensee found red iron oxides, indicative of corroding carbon steel.
~
Reason for Reporting to IRS In addition to the' integrity of many safety-related systems being in jeopardy due to the. loss of the salt water cooling system, this event highlighted the significance of a potentially serious problem which may result from contaminated control air systems. This falls into category 4, "Significant Generic Problems."
Actions Taken
~
1.
The licensee has undertaken a major o erhaul of the plant's preventive maintenance program. The licensee hired an engineering consultant-to prepare a' detailed computerized maintenance program.
2.
The olant's entire instrument air system has been blown down.
New desiccant has been installed, as has a new air filtration system including instrumentation to measure the pressure drop across the filtarc.
The licensae's new,--
improved, preventive maintenance program will address the condition of the desiccant.
3.
The in-service testing program has been upgraded.
Salt water cooling pump testing now includes vibration measurement.
4.
The auxiliary SWC pump's priming system has been modified in an effort to improve its reliability.
Furthermore, the licensee is planning to include the auxiliary SCW pump in their in-service' testing program after all. proposed modifications are complete.
' OECD NUCLEAR ENERGY T
g D
AGENCY
{
n 38, bd.Suchet75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telex 630668 AEN/NEA RESTRICTED NO. IRS DIFFUSION RESTREINTE Titre Title Foreign Material in Air System Blocked Solenoid Exhaust Country - Pays Date of Incident - Date de l' incident January 25. 1982 United States Type of Reactor - Type de tsacteur PWR Plant - Centrale Licensee - D6tenteur du permis d'exploitatior Ginna Rochester Gas and Electric
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Unit N*
- Tranche n*
Westinghouse Puissance First Commercial Operation -
Power 490
'MWe(net)
Date de mise en service pg7cy 3979 Systems or Components Affected - Systkmes ou composants affectds Solenoid Valve Initial Plant Condition - Etat initial de la tranche Emergency Shutdown Way in which Incident was Detected ?
Comment l' incident a-t-it'6td ddtect6 ?
Control Board Indication Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou lib 6tation de radioactivit6 Date of Receipt - Date de r6ception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projet6es et w.w wx w__ri -.,, -,.
o Event Description On January 25, 1982, while opening the pressurizer power-operated relief valve
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(PORV) the fourth time in an effort to reduce primary and secondary pressure below the main steam safety valve setting during a steam generator tube rupture incident, the PORV stuck open.
Cause of Event l
In order to meet the PORV function of low temperature overpressure protection',
I the vent side of the actuating solenoid valve was restricted. The solenoid valve was replaced in 1981 with an environmentally qualified ASCO NP-1 which cannot have a restricted vent.
Evidently foreign material in the air system blocked the exhaust in the solenoid since there is no air filter in the system for the solenoid.
Reason for Reporting to IRS l
This event is another example of a possible problem with contaminated air control systems.
It falls into category 4, "Significant Generic Problems."
Actions Taken Reactor coolant system depressurization was stopped by. closing the block valve.
The restriction was removed and an orificed check valve was installed upstream of the PORV pneumatic operator.
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'OECD NUCLEAR ENERGY T
A o
AGENCY
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38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM TeI,524.96.93 Teixx 630668 AEN/NEA RESTRICTED NO. IRS DIFFUSION RESTREINTE Title - Titre Failure of Diesel Generator Start Test Due to Contaminated Air Start System Country - Pays Date of Incident - Date de l' incident June 20,1981 United States Type of Reactor - Type de tsacteur BWR Plant - Centrale Licensee - Dstenteur du permis d ' exploitation Browns Ferry Tennessee Valley Authority
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Tranche n*
1 Unit N*
General Electr'ic Power - Puissance First Commercial Operation -
August 1974 l
1067 MWe(net)
Date de mise en service composants affectss Systems or Components Affected - Systhmes ou Diesel Generator, Diesel Generator Air Start System Initial Plant Condition - Etat initial de la tranche Refueling Outage Way in which Incident was Detected ?
Comment '1' incident a-t 'il des dstects ?
Surveillance Test Radiation Exposure or Radioactivity Release -
1 Exposition aux rayonnements ou libdration de radioactiviti l
None Date of Receipt - Date de rsception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned
(
(safety significance of incident) should be included in the following pages.
Descrir9 tion de l' incident, causes possibles, mesures prises ou projetnes et
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8 Event Description During diesel generator redundant start test, left bank starters would not develop minimum rotational speed. The diesel was removed from service to repair the air start motors.
Cause of Event Nonnal wear, dirt, and grit in the air start system.
Reason For Reporting to IRS This event falls into category 4, "Significant Generic Problem" because it represents a continuing problem of contaminated air systems.
Actions Taken The air start motor was replaced. To prevent recurrence, all air start motors and actuation valves will be inspected s mi-annually and all air start motors will be rebuilt annually.
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.' OECD NUCLEAR ENERGY
{T D
AGENCY
.~
38,bd.Suchet75016 Paris INCIDENT REPORTING SYSTEM Te1. 524.96.93 Telex 630 668 AEN/NEA RESTRICTED l
NO.lRS DIFFUSION RESTREINTE l
Title - Titre Degradation of Performance of Sample Line Isolation Valve Due to Contaminated Control Air s
Country - Pays Date of Incident - Date de l' incident July 7, 1981 United States Type de r6acteur Type of Reactor PWR Plant - Centrale Licensee - D6tenteur du permis d* exploitation Rancho Seco Sacramento Utility District
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Tranche n*
1 Unit N*
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Babcock i Wilcox Power - Puissance First Commercial Operation -
913 nwe(net)
Date de mise en service April 1975 Systems or Components Affected - Systemes ou composants affectss Pneumatically Operated Globe Valve Initial Plant Condition - Etat initial de la tranche l
l 100% Power Way in which Incident was Detected ?
Comment l' incident a-t-il,dtd d4tects ?
Surveillance Test Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou lib 6 ration de radioactivits None Date de r6ception Date of Receipt Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages, Description de l' incident, causes possibles, mesures prises ou projetses et me ha a0lemRfD doivent fipurer
Event Description On July 7,1981, during surveillance testing of the once through steam generator, a sample line isolation valye exceeded the allowable time limit for closure.
Cause of Event Foreign material, which appeared to be desiccant from air dryers, was found in the air discharger part.
Reason for Reporting to IRS This event, although minor in itself, is another example of a possible problem with contaminated control air systems.
It falls into category 4, "Significant Generic Problems."
Actions Taken The air discharge port was cleaned and the valve operated properly. A new air dryer with a larger capacity has been installed in the instrument air system. The filters in this device will be changed every 90 days and an inline dew point monitor has been installed. Such surveillance should assure that such a problem does not recur.
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OECD NUCLEAR ENERGY AGENCY 38, bd.Suchet75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telex 630668 AEN/NEA RESTRICTED No.lRS DIFFUSION RESTREINTE Title - Titre Failure of Gas Turbine Generator to Start due to Contaminated Control Air.
Country - Pays Date of Incident - Date de l' incident Julv 14.1981 United States Type of Reactor - Type de tsacteur BWR Plant - Centrale Licensee - Dstenteur du permis d* exploitation l
Northeast Utilities Millstone
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Unit N*
- Tranche n *
}
Power - Puissance First Commercial Operation -
652 nwe(iset)
Date de mise en service 12/70 Systems or Components Affected - systames ou composants affectss
. Gas Turbine, Generator, Contro1 Air Initial Plant Condition - Etat initial de la tranche Power increasing from 85%
Way in which Incident was Detected ?
Comment l' incident a-t-il sts ddtects ?
I Gas Turbine Generator Failed to Start during Testing Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou libsration de radioactivits None Date of Receipt - Date de rsception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages, ibles, mesures 19 rises ou T9rojetees et
o Event Description On July 14, 1981, at 1425 hours0.0165 days <br />0.396 hours <br />0.00236 weeks <br />5.422125e-4 months <br /> while testing the Gas Turbine for Unresolved Safety Issue A-44 " Station Blackouts.", the Gas Turbine failed to accelerate upon receipt of a Black Start signal. The Black Start signal for the Gas Turbine was initiated from the control room. The Gas Turbine followed its planned sequence of start up, to the point where the start relays picked up.
The " start signal received", and " sequence in process" indicators illuminated. The next step in this sequence never occurred.
Cause of Event Investigation into the incident revealed that the shut off valve for the air start motor stuck in a closed position. This valve is one of two that functions to admit air to the air start motor, which in turn accelerates the gas turbine to light off speed.
The source of the contaminant is the starting air compressor receiver which is fabricated of steel and has a large internal surface area that is susceptible to rusting from condensation.. Investigation into the type of. paint to paint the internals of the air receiver to eliminate the rust particles that settle in the valve is in process.
In addition installing filters in the air lines is falso a consideration.
Reason for Reporting to IRS This event falls into two areas:. category 2
. Loss of An Essential Support System" and category 4 "Significant Generic Problems." Events caused by con-taminated air systems have also occurred at other plants.
l Action Taken The valve was removed from the engine for disassembly to determine the cause.
of-its malfunctioning.
Inspection of the internals showed accumulations' of rust. The valve and valye operator internals were thoroughly cleaned, and teflon seats in the ball valve were replaced. The valve was bench tested and i
reinstalled.
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OECD NUCLEAR ENERGY g
D AGENCY s
38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telax 630668 AEN/NEA RESTRICTED N2. IRS DIFFUSION RESTREINTE Title - Titre Contaminated Air Systems Country - Pays Date of Incident - Date de l' incident United States Numerous Type of Reactor - Type de r6acteur Plant - Centrale Licensee - Dstenteur du permis d' exploitation Many Unit N*
- Tranche n *
" ""' ##"{'* ~ I A'"#
Puissance First Commercial Operation -
Power MWe(net)
Date de mise en service Systems or Components Affected - Systhmes ou composants affectis Initial Plant Condition - Etat initial de la tranche Way in which Incident was Detected ?
Comment l' incident a-t-il; sti ddtect4 ?
Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou lib 6 ration de radioactiviti Date of Receipt - Date de rdception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetnes et i
m-m _ s uam _ n _ m,a _ m ru__
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IRS Report on Contaminated Air Systems An increasing number of operatin'g events involving contaminated air systems have been reported. The attached events are grouped together as one report because they are indicative of a potential generic concern and not necessarily significant enough to be the subject of individual reports. Additional cmpressed air problems have been discussed in " Compressed Air and Backup N,itrogen Systens in Nuclear Power Plants" (0RNL/NSIC-206).
Attached Reports:
1.
Failure of Gas Turbine Generator to Start Due to Contaminated Control Air.
2.
Degradation of' Performance of Sample Line Isolation Valve Due j~
to Contaminated Control Air.
3.
Failure of Diesel Generator Start Test Due to Contaminated Air Start System.
4.
Foreign Material in Air System Blocked Solenoid Exhaust, i
5.- Failed-Air Supply Solenoid 0-Ring Disables North Salt Water-i Cooling Pump.
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,OkD NUCLEAR ENERGY T
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. AGENCY
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. 3 8, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Tela 630668 AEN/NEA RESTRICTED N2.lRS DIFFUSION RESTR.EINTE Title - Titre Valve Overtravel Anomalies Country - Pays Date of Incident - Date de l' incident United States Type of Reactor - Type de r4acteur PWR Plant - Centrale Licensee - Ditenteur du permis d' exploitation Millstone Northeast Utilities
"* ~
Unit N*
- Tranche n
- 2 Combustion Engineering Power - Puissance
.First Co*mm er ci a l Operation -
870
""'I"'*I D'**
d' *i * '" **i
12/75
' Systems or Components Affected - Syst mes ou composants affectds High Pressure Safety Injection (HPSI) valves Initial Plant Condition - Etat initial de la t,r a n ch e 100% Power Way in which Incident was Detected ?
Comment l incident a-t-il dtd ddtects ?
Routine Surveillance Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou libdration de radioactiviti None Date of Receipt - Date de rdception Date of' Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned
}
(safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetnes et enseignements tirss (signification de l' incident pour la s0rets) doivent figurer les paces suivantes.
sur
.s Event nescription While performing surveillance testing between June 16 and July 9, 7 of the 8 HPSI motor-operated valves overtraveled from 1/4 to 1 inch past the required position. Overtravelling results in higher injection flow rates and smaller HPSI pump Funout margins.
Cause The overtravel was caused by the variable coastdown of the Limitorque motor-operated valves and the small margin of accept'able valve position.
Limitorque Adjustment Manual implies that the " fully open" limit switch can be activated from 90 to 100% of the valve's full open position. The desired valve opening necessary to get acceptable HPSI flow for the injection valves at Millstone 2 is 2h turns of a " fully open" limit switch setting of about 71% of valve stem travel.
Limit switch adjust-ments in this vicinty are somewhat more difficult to set than at the recommended 90-100% position.
Further, all Limitorque actuators are designed to reach full motor speed before applying torque to turn the valve stem. This results in an impact to the valve stem at energization that'should result in less precise initial steam movement.
In addition, the motor speed during valve movement is not constant, but is load ds-pendent and may be influenced by the relative valve packing tightness
_ _ on the test date.
Also, there is a certain amount of play or lost motion in the worn gear linkage to the limit switch.
Finally, the Limitorque operator is relatively large in comparison to the HPSI valve it is driving.
Each of these conditions further exacerbate the diffi-culty in reaching repeatable valve positions while trying to stop valve movement in a narrow band at about 70% of stem travel.
Reason for Reporting Because of the possibility that there may have been an interface problem between Combustion Engineering-(CE) and the architects / engineers during-plant construction, this occurrence is reportable under criterion 3 "Significant Deficiencies in Design Construction, Operation or Safety Evaluation."
Actions Taken To eliminate the potential for future overtravel events the valves are opened to the required position and maintained at that position.
Other possible corrections could be either installation of atsmaller" flow restrictive device, or the use of a larger plug,or reduction 6f the trim on the HPSI valves. These measures should be taken to ensure that the rated HPSI flow is achieved at the' full open HPSI valve position.
' '05CD NUCLEAR ENERGY T{
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9 AGENCY 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telax 630668 AEN/NEA RESTRICTED Nr.lRS DIFFUSION RESTREINTE Title - Titre Inadvertent Containment Spray Actuation Country - Pays Date of Incident - Date de l' incident February 11, 1981 United States Type of Reactor - Type de r6acteur PWR Plant - Centrale Licensee - Datenteur du permis d exploitation Sequoyah -
Tennessee Valley Authority
' ~
Unit N*
- Tranche n*
}
Westinghouse Power - Puissance First Commercial Operation -
7jg3 MWe(net)
Date de mise en service 3349 composants affect 4s Systems or Components Affected - Systhmes ou Reactor Coolant Pressure Boundary, Residual Heat Removal System, Containment Spray System Initial Plant Condition - Etat initial de la tranche Cold Shutdown Way in which Incident was Detected ?
Comment l' incident a-t-il sts dstect6 ?
Notice Decreasing Pressurizer Pressure and Level, Operator Error l
Radiation Exposure or Radioactivity Release -
Exposition aux rayonnements ou lib 6tation de radioactivits I
None i
Date of Receiot - Date de r6ceotion Date of Distribution - Date de distribution 1
l l
Event description, possible causes, actions taken or planned and lessons learned l
(safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projet6es et
[
enseignements tirss (signification de l' incident pour la s0rets) doivent figurer
i Event Description i
On February 11, 1981, while in cold shutdown, Unit 1 experienced a loss of coolant event. The containment spray system was inadvertently activated for 35 minutes and released approximately 105,000 gallons of water into the containment area.
Cause During a shift change the oncoming uni't op'erator (U0) was informed of the need to confirm that the residual heat removal (RHR) "A" containment spray valve was fully closed and that the RHR "B" train should be returned to service. The UO then instructed an auxiliary unit operator (AU0) to manually close the motor-operated spray valve and open two manually-operated RHR crossover valves. The AU0 opened the two crossover valves and erroneously opened the spray valve.
Adequate training did not appear to be provided for the AU0 before being permitted to assume his duties.
Also, inadequate oral communication appeared to exist between a licensed and non-licensed operator.
Reason for Reporting This event is considered to be reportable pursuant to criteria 2.2, since'the reactor coolant pressure boundary was breached, and criteria 3, since the event occurred due to a deficiency in operation.
Actions Taken In response to decreasing pressurizer level and pressure, the control room operator opened the suction valves from the refueling water storage tank to both RHR pumps and tripped the two operating reactor coolant pumps.
The operator did not close the RHR suction to the hot leg of the reactor coolant system or the RHR connection to letdown to the chemical and volume control system.
Numerous personnel errors can be attributed to inadequate or misunderstanding of instructions during recent nuclear plant events. To aid in diminishing these operator deficiencies the licensee has taken the following long-term actions:
1.
Upgraded the formal training program to include on-the-job training for AU0s.
2.
Established administrative procedures for verifying oral communications between operators.
3.
Establishing procedures to mitigate a loss-of-coolant event while in cold shutdown.
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OEED NUCLEAR ENERGY T
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AGENCY i
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. 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Tel:x 630668 AEN/NEA RESTRICTED N D. IRS DIFFUSION RESTREINTE Title - Titre Loss of 125 V DC Bus.
Country - Pays Date of Incident - Date de l' incident United States
"~Y Type of Reactor - Type de re,acteur P,WR i
Plant - Centrale Licensee - Ddtenteur du permis d' exploitation Millstone Northeast Utilities
' " ~
Unit N'
- Tranche n*
g Combustion Fnginparing Bower - Puissance
.First Co*mmercial Operation -
MWe(net)
Date de mise en service
$gj75
' Systems or Components Affected - Systames ou composants affectds DC Power System Initial Plant Condition - Etat initial de la t,r a n ch e 100% power.
Way in which Incident was Detected ?
Comment 1 *~inciden t a-t-il std ddtects ?
Operator observation Raciation Exposure or Radioactivity Release -
Exposition aux rayonnements ou libdration de radioactivits None Date of Receipt - Date de rdception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
{
Description de l' incident, causes possibles, mesures prises ou projetses et l
enseignements tirss (signification de l' incident pour la sGrets) doivent figurer j
sur les pages suivantes.
6 Event Description The event was initiated by a plant equipment operator while perform-ing his shift duties.
Instead of operating a 125 V de system ground detector switch, he mistakenly used the adjacent control switch and opened breaker D0103 which de-energized the "A" facility 125 V dc bus 201 A (see Figure 1).
Loss of the.de bus caused the opening of four (out of a total eight) reactor trip circuit breakers (TC8s) which tripped the reactor.
The loss of the de bus also caused complete loss of all control room annunciators and started the "A" facility diesel generator. The operators noticed the trip by observing Control Element Assembly (EA) inward motion (the four TCBs'that opened were without indication because of loss of de voltage, and the other fodr were indicated closed).
The reactor trip should have caused a turbine trip and consequent generator lockout.
However, since the trip relay in the turbine Electro-Hydraulic Control (EHC) system receives its power from de bus 201A, the automatic turbine trip did not occur. Approximately 29 seconds after the loss of the de bus, the control room operator operated the master trip button on the EHC system panel. This caused a turbine trip, since the 24 V de power required was available from the permanent magnet generator on the turbine shaft. The turbine stop valve closed, but the generator. remained tied to the 345 KV system.
On a turbine trip, the station auxiliary loads would normally fast-f transfer from the normal station service transformer (NSST) to the reserve station service transformer (RSST) (see Figure 2).
During the event, however, only relay 94TG, which is connected to the "B" facility 125 V de system, operated on the turbine trip from the i
24 V de logic.
Relays 83X1, and 83X2, which are connected to the "A" facility 125 V dc, did not operate. Contacts of the 94TG relay normally trip the four NSST breakers (two 6900 V bus breakers and two p
4160 v bus breakers).
The two breakers whose control power is supplied by the "B" facility dc bus tripped and de-energized 6900 V bus 25B and 4160 V buses 24B and 24D. The two breakers whose control power is supplied by the "A" I
facility de bus did not trip since no control power was available.
y Therefore, buses 25A (6900 V) and 24A and 24C (4160 V) remained con-nected to the main generator and the 345 KV switchyard through the NSST.
f l
The main generator breakers 9T2 and 8T2 in the switchyard remained f
closed because trip lockout relay 94MGI is also connected to the "A" j
facility 125 V de bus.
Since ralavs 83X1 and 83X2 did not operate (see Figure 3), the RSST l
l supply breakers for buses 24D and 25B did not receive a close signal.
[
With 4160 V bus 24D remaining de-energized, the emergency safeguards actuation system (ESAS) detected a loss of normal power (LNP), which t
r
~ caused load shedding and isolation of bus 240. The "B" emergency diesel (13U) automatically started and energized that bus.
Buses 248 and 25B remained de-energized. The loss of the "A" de bus also caused a " black start" of the "A" emergency diesel generator (12V), since the air supply solenoid valves that supply engine starting air are designed to open upon loss of power (i.e., de-energized to open). Since de power was not available, no field flashing occurred and output breakers remained inoperable.
Approximately 51 seconds following the initiation of the event, the "A" de bus was re-energized.
NSST supply breakers of buses 24A and 25A tripped and the buses fast-transferred to the RSST on operation of relay 83X1 and the fast transfer logic.
Relay 83X2 operation caused RSST supply breaker of 6900 V bus 25B to clos'e attempting to energize all its connected motors (two reactor coolant pumps -
and one condensate pump) - this caused the breaker to trip on overcurrent.
Relay 94MGI also operated and tripped the generator output breakers in the switchyard. The main steam isolation valves closed upon re-energization of the de bus due to the control schemes of the air supply and vent solenoid valves in the valve actuator.
Closure i
of Main Steam Isolation Valves (MSIVs) caused the tripping of both main feed pumps. Auxiliary feed pumps were manually started, supplying both steam generators.
"A" emergency diesel generator (EDG) failure-to-start relay (SFR) energized upon re-energization of the de bus. This in turn ener-gized the diesel shutdown relay (SDR) which stopped the engine (See I
Figure 4). On loss of de control power and subsequent re-energization of the de bus the SDR could be energized by operation of either the loss of.dc power relays (CR-1 & CR-2) or the failure-to start relay.. Once tripped, local reset action would be required to restart the diesel engine.
About nine minutes into the event, "B" EDG tripped and de-energized bus 24D. The EDG was declared inoperable due to a service water leak that sprayed the machine. After reviewing available data, the licensee concluded that the salt water spray had caused malfunction of the governor control, which had caused the engine to run down in speed. The two-out-of three low lube oil pr, essure trip had consequently tripped the engine.
(
It was also concluded that the loss of several instrumentation power supplies connected to the diesel generator bus was caused by the operation of the input fuses in the transformer of the power supplies. This was due to the low frequency caused by the slowing of the diesel generator. The loss of the power supplies caused the loss of several indicators in the control room.
On loss of the "B" EDG, the control room operator connected bus 24D to the RSST after overriding the undervoltage signal.
It was noted at this time that several indications were not available (including auxiliary feedwater flow to the No.1 SG, main steam header pressure,
make-up tank level, charging pump pressure, etc).
One to two hours later, all fuses were replaced and the instrumentation restored.
Loss of this instrumentation hampered operator actions during the recovery operation.
One-half hour following the reactor trip, the pressurizer level re-turned to 40%, its programmed value for hot zero power. The pressurizer pressure was restored to 2250 psia 45 minutes following the reactor trip by the pressurizer heaters. While in the hot shutdown condition, the atmospheric dump valves (ADVs) were automatically controlling reactor coolant average temperature to 541*F, nine degrees above the normal zero power Tave of 532*F and the operator was trying to control pressurizer pressure by spraying with the nomal spray system. However, only two reactor coolant pumps, P40A and P40C, were operating because the RSST supply breaker for the 6900 V bus 258 had tripped on over-current. The operators assumed this two-pump combination provided effective pressurizer spray since, as shown in Figure 5, one of the two lines to the common spray header comes from the cold leg at the P40A reactor coolant pump discharge.
However, the operators were having difficulty controlling the pressure and therefore assumed non-condensables were present in the pressurizer.
Subsequent review of initial plant startup test data has shown that this two-pump com-bination results in no significant pressurizer spray flow.
Reactor coolant average temperature decreased when the ADVs opened, resulting in a decrease in pressurizer level and pressure.
The decrease is pressurizer pressure caused the liquid to flash resulting in an increase in the steam mass. When the ADVs closed, the reactor coolant average temperature and the pressurizer level increased, compressing a larger steam mass.
Since the two cycles of opening and closing of the ADVs immediately preceeding the lifting of the Power Operated Relief Valves (PORVs) caused the pressurizer pressure and level to drop a successively' lower value, the mass of steam increased with each cycle. Due to the quiescent liquid surface, the heat transfer from the steam to the pressurizer liquid was small.
In addition, since the spray flow was inadequate and the heat transfer to the wall is small, minimal condensation occurred.
The increase in pressurizer level caused the steam mass to be compressed nearly isentropically, which resulted in an increase in pressure and temperature of the steam to become superheated rather than condense, leading the operators to believe that the abnormal pressure response was caused by the presence of non-conden-l sable gases in the pressurizer.
Since the cycling of the ADVs caused the mass of steam to increase.the pressurizer level swings caused the pressure to reach higher peaks in each successive cycle. When the pressure increased to 2380 psi, approximately two hours and 15 minutes after the reactor trip, the PORVs opened briefly.
The operators then began to use the l
auxiliary spray and the pressurizer pressure was successfully lowered.
l A plant cooldown to cold shutdown condition was then begun.
During the operation of the PORVs the temperature in the discharge line and the quench tank parameters increased as expected.
However, the acoustic valve monitoring system (AVMS) did not respond.
Subsequent l
. verification by the licensee (by manually opening a PORV) indicated the AVMS was operable.
Since the AVMS is designed with a 2.4 second time delay, it was concluded by the licensee that AVMS did not actuate because of the extremely short duration that the relief valves were open.
Cause of Event The cause of the accident was operator error.
Reason for Reporting This event involved several related and unrelated incidents:
(1)
Partial loss of normal offsite power, (2)
Complete loss of control room annunciators (3)
Inoperability of both energency diesel generators (4) Loss of several indications in the control room, and (5)
Ineffective pressurizer spray through the normal spray system.
This event is being reported under criteria 2.6 " Loss' of an Essential Support System" and criteria 3 "Significant Deficiencies in Design, Construction, Operation, or Safety Evaluation."
Actions Taken The licensee plans to provide permanent breaker identification labels and review the plant equipment operators rounds to identify similar l
situations.
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OECD NUCLEAR ENERGY AGENCY 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telu 630 668 AEN/NEA RESTRICTED NC. IRS DIFFUSION RESTREINTE Title - Titre Corrosion of Studs on Valve in Spent Fuel Pool Country - Pays Date of Incident - Date de l 'inc iden t February 17, 1982 United States Type of Reactor - Type de tsacteuz PWR Plant - Centrale Licensee - Dstenteur du permis d'exploitatior Maine Yankee Atomic Power Company Maine Yankee
~
Unit N*
- Tranche n*
Combustion Engineering Power - Puissance First Commercial Operation -
g NWe(net)
Date de mise en service composants affectds Components Affected - Systhmes ou Systems or Spent Euel Cooling Pump Discharge Valve Initial Plant Condition - Etat initial de la tranche 97% power Way in which Incident was Detected ?
Comment l' incident a-t-ib sts ddtects ?
Valve repair Radiation Exposure or Racioactivity Release -
Exposition' aux rayonnements ou lib 6 ration de radioactiviti Date of Receipt - Date de rdception Date de distribution Date of Distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
TQagsgip; tion de l' incident, causes possibles, mesures prises ou projetses et
Event Description During normal operation, while disassembling a spent fuel pit cooling pump discharge valve, two 5/8-inch studs were brokan in the disassembly process and found to be corroded.
The spent fuel pit cooling water was isolated.
I Cause of Event lhe broken studs were manufactured using ASTM A-93, grade B6, Type 416 sta.inless steel.
It has been concluded that this material is perhaps not well suited for systems that carry borated water.
Reason for Reporting This event falls under categories 3 and 4, "Significant beficiencies in Design, Construction, Operation or Safety Evaluation" and "Significant Generic Problems" respectively.
Actions Taken Maine Yankee is initiating a program to eventually replace all of the Type 416 bolts 'io valves that contact borated water.
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38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Tel:x 630668 AENnJEA RESTRICTED Nr.lRS DIFFUSION RESTREINTE Title - Titre Crack Indications on Steam Generator Primary Manway Studs Country - Pays Date of Incident - Date de l' incident United States July 22. 1980 Type of Reactor - Type de rdacteur PWR Ddtenteur du permis d ' exploitation Plant - Centrale Licensee Arkansas Nuclear One Arkansas Power and Light Company Manufacturer - Fabricant Unit N*
- Tranche n, 3
ggggggy.and Milcox Power - Puissance First Commercial Operation -
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12/74 836 composants affectds Systems or Components Affected - Systsmes ou Steam Generator Manway Studs Initial Plant Condition - Etat initial de la tranche Shutdown Way in which Incident was Detected ?
Comment l' incident a-t'll 4td ddtects ?
Nondestructive testing Radiation Expcsure or Radioact2Vity Rclease -
libdraticn de radiorctivits Exposition a u; rayonnements ou Date de rdception Date of Receipt Date of Distribution - Date de distribut; c.n Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Qaac2fug_i_on de__l'_inc_ide_nt o causes possibles, mesures prises ou projetses et
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Event Description During a steam generator tube leak unit outage, non-destructive testing was conducted on the upper and lower once through steam generator (OTSG) primary manway studs prior to reinstallation.
The inspection revealed three lower manway studs with indications of cracking on the "A" 0TSG. No indications were found on the "B" 0TSG.
Cause Analysis of one of the studs indicated that the ' crack initiated and propagated by a stress corrosion mechanism.
Further analysis identified the presence of molybdenum oxide, sulfur as a sulfide with a small amount of sulfate, boron, and lithium.
Reason for Reporting to the IRS Due to the increased incidence of cracking of manway studs because of corrosion, this is reportable pursuant to criteria 4 "Significant Generic -Problems" as well as criteria 2, "Significant Degradation of Safety-Related Systems."
Action Tak,qn The three studs with crack indications we,re replaced. The OTSG manway studs will be reinspected periodically as required by the established In-service Inspection Program.
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AGENCY 38, bd. 5uchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telex 630668 AEN/NEA RESTRICTED No. IRS DIFFUSION RESTREINTE Title - Titre Core Barrel Assembly Thermal Shield Bolts Broken Country - Pays Date of Incident - Date de l' incident United States July 15.1981 Type of Reactor - Type de r6acteur PWR Plant - Centrale Licensee - D6tenteur du permis d*exploitatior Duke Power Company Oconee Unit 1
" ""' '*" ~ ' b'A' "*
unit it * - Tranche n*
Babcock and Wilcox Power - Puissance First Commercial Operation -
g59 MWe(net)
Date de mise en service 393,3973 Systsmes ou composants affectEs Systems or Components Affected Four thermal shield bolts, one guide block, four bolt locking cups (three missing, one partially attached)
Initial Plant Condition - Etat initial de la tranche Defueled Way in which Incident was Detected ?
Comment l' incident a-t-il 6t4 ddtects ?
Ten year inservice inspection Radiation Exposure or Raaicactivity Release -
Exposition aux rayonnements ou libdration de radioactivits Date of Receipt - Date de r6ception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Descrirgtion de l' incident, causes possibles, :nesures prises ou projet6es et
Core Barrel Assembly Thermal Shield Bolts Broken at Oconee Unit 1 Event Description During the visual examination of the reactor vessel internal components, unexpected conditions were observed. The. following table summarizes the results of the initial visual examindtion:
1.
Four of 96 bolts connecting the thermal shield to the lower grid flow distributor flange were missing.
2.
Approximately 80% of the remaining thermal shield bolts were backed out from 0.1 to 0.5 inches.
3.
Three bolt locking cups were missing.
4.
One locking cup partially attached.
5.
One guide block on the Y-axis was missing.
The following table summarizes the current status of components missing and those retrieved at the bottom of the reactor vessel:
Weight Initially Still (lbs) Dimensions Missing Retrieved Missing Guide Block 18.0 3"x6.5"x5" 1
0 1
Guide Block Dowel 2.3 4.5", 1.5"D 1
0 1
Guide Block Bolt 0.902 4.1"x 1.7"D, 1.0D 1
0 1
Guide Block Bolt Washer 0.085 2" 00, 1.0 ID 1
0 1
Thermal Shield Bolt Heads 0.582 1.375"x 1.75D 5
3 2
Thermal Shield Bolt Shanks 0.669 5.125 1.00 4
4 0
Thermal Shield Locking Clips 0.124 1.0"x2.5" x 1.75" 3
1 2
The retrieved components have been shipped to B&W, Lynchburg, for complete examination and determination of cause of failure.
+.
The visual examination of the selected areas has revealed no other significant deficiencies.
The following table summarizes the inspection results:
Thermal shield to lower grid joint No distress of metal Upper thermal shield restraint Locking clips intact; no visual evidence of wear Core guide blocks Welds intact; indication of guide block and lug contact Flow Distributor, outside No indication of impact damage
. Incore instrument guide tubes No indication of impact damage RV guide lugs Some indication of contact Cause of Event An apparent cause has not been determined for this event, although a preliminary evaluation has.been made of the safety implications of the observed conditions. This safety evaluation considered'the following:
1.
Structural implications of the thermal shield bolt failures 2.
Structural implications of the guide block failures 3.
Loose part implications, i.e., damage to the fuel, interference with CRD motion and damage to other RCS canponents due to loose parts.
Due to the function served by the thermal shield and the manner in which it is structurally considered in the accident analyses, the observed conditions are not believed to have significant public health and safety implication.
Each of the above three types of safety implications is discussed in detail
~
below.
1.
Thermal Shield Bolts The thermal shield is not a principal load carrying member of the reactor internals, i.e., its function is to reduce radiation effects on the reactor vessel.
In spite of this function, however, several consequences of joint degradation were considered at the upper and lower end of the thermal shield.
If the upper restraint becomes loose, the thermal shield response due to fluid loadings will change with the most likely consequences being a reduction in natural frequency of the shield. This could lead to a significant increase in the cyclic stresses of the lower and attachment bolts. As looseness at the upper restraint develops, any significant metal-to-metal impact would be most likely detected by the loose parts monitoring system (LPMS).
Detection becomes increasingly probable at higher frequencies. Should the lower attachment bolts fail, the shrink.fjt between the lower grid flange and the thermal shield could then loosen and vertical motion would be possible.
In the upward direction, motion would be limited by the core barrel flange and stop.
In the downward direction motion is limited since the thermal shield rests on the lower grid flange. Therefore, vertical motion is constrained in both directions but should significant vertical motion occur, metal-to-metal impact would also occur and the LPMS would indicate the condition before serious damage would occur.
Before vertical motion and associated impacting could occur, numerous loose parts (i.e., bolts, locking cups, etc.) would also exist in the system and again the probability of detection by the LPMS is high.
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. Although not considered credible, the extreme condition considered was complete failure of the lower grid flange to which the thermal shield is attached. Even under this extreme condition, the core support assembly would remain intact but the thermal shield.could conceivably drop a short distance and then be restrained by the twelve core support lugs. These core support lugs are designed to accommodate the design weight of the core and thermal shield, which together, are 13 times the weight of the thermal shield alone. The failure of the lower grid flange is considered to be an extremely remote possibility but nevertheless one in which core cooling would be unaffected.
In summary, evaluation of failure consequences considerably more severe than those observed are not considered to represent a significant risk to public health and safety because of the purpose served by the thermal shield and the lack of adverse effect on core cooling.
2.
Guide Blocks The guide blocks are attached to the lower RV internals and in the original design they were to provide. lateral (side) restraint for seismic loadings.
During recent analyses however, including the analysis of the effects of LOCA-induced asymmetric forces, no restraint was assumed at the bottom of the core support assembly and all stresses were found to be within ASME code allowables. Therefore, the guide b, locks are not essential to assuring the integrity of the reactor internals under accident loads.
Furthermore, it appears that the guide block failure.is independent of the thermal shield bolt failures and would seem to be an isolated event based on the normal appearance of the dowel pins in the other.23 guide blocks. The single guide block failure appears to be an isolated event but even if this were not the case additional failures would not have significant safety consequences aside from the loose parts implications which are addressed below.
3.
Loose Parts j
The size of the loose parts which have resulted from these failures vary widely from the locking clip or a fraction thereof to the guide block.
(There is some question at this time as to whether the guide block might have been missing when the internals were last installed in 1976.)l which are larger than theAny loose parts in the head lower internals region of the reactor vesse flow passages in the fuel assembly end fittings would be precluded from passing l
through the core or enter,ing the remainder of the reactor coolant system.
Pieces which are small enough to pass through the fuel assemblies and into the reactor coolant system are not large enough to seriously degrade the RCS pressure boundary with the possible exception of the steam generator tubing or tube to tube sheet joint.
Impacts on the generator upper tube sheet from an object as small as 1.3 ounce, has been detected by the Loose Parts Monitoring System.
Even if not detected, however, the most significant consequences would be primary to secondary leakage which is detectable and would not interfere with an orderly shutdown.
1
. In no case is it anticipated that fuel damage would occur due to either mechanical efforts or flow blockage. This is because pieces which are small enough to pass through the fuel assembly end fitting would be expected to pass,on through the core, and out of the reactor vessel.
Should a small piece lodge in a fuel assembly grid spacer, the effect would be quite localized and could conceivably cause localized fuel damage. Any fuel damage great enough to breach the cladding would be readily detected.
The remote possibility also exists that a larger piece could cause some flow blockage in the lower grid area but because the lower end of the active core operates at reduced heat rates, no fuel damage would be anticipated.
The possible effects of loose parts were considered in connection with inter-ference between control rod pins and guide tubes. This is not considered likely because of the small diameter (1/8") coolant entry at the lower end of each guide tube. This would require not only a very small piece but also a precise flow direction to enter the guide tube.
Furthermore, the velocity in the guide tube, immediately past the entrance decreases significantly so that a metallic object is not likely to be supported by the vertical fluid str' sam.
However, although control pin interference is considered very improbable, if it were. assumed to occur, it would very likely be detected during control rod exercise programs. This is not considered to be a problem because any pieces small enough to reach the upper plenum area would not be expected to lodge between a control pin and guide tube but rather pass on through the upper plenum.
If a loose part were to reside in the lower plenum of the reactor vessel, damage to the incore guide tubes or incore nozzles could occur if the part were located in a highly turbulent area.
These, however, are not pressure boundary parts.
Furthermore, repeated impacts from a loose part (approximately a 2 pound RC l
pump impeller nut) have been detected in the past by the LPMS.
Somewhat smaller parts than the pump impeller nut should also be detectable in this area.
In summary the effects of loose parts in the reactor coolant system do not represent a threat to public safety.
Experiences in several operating reactors have proven this to be the case.
l l
Reason for Reportina to IRS This occurrence is considered to be reportable pursuant to criterion 3, "Significant Deficiencies in Design, Construction, Operation or Safety Evaluation" and 4, "Significant Generic Problems."
Actions Taken 1.
Evaluate the loose parts monitoring system and implement hardware / procedural changes as determined necessary.
Sensitivity checks of the Unit 1 LPMS are completed and the Unit 2 and 3 LPMS have been recalibrated.
2.
Determine and implementinspection plans for Oconee Units 2, 3 vessel internals, as appropriate.
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. 3.
Determine and implement plan for additional inspection of Oconee Unit 1 internals.
4.
Develop and implement plan to remove Unit 1 thermal shield bolts.
5.
Examine failed bolts and determine cause of failure.
6.
Evaluate alternative design concepts a'd implement selected design.
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OECD NUCLEAR ENERGY
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AGENCY
>s 38, bd. Suchet 75016 Paris INCIDENT REPORTING SYSTEM Tel. 524.96.93 Telex 630668 AEN/NEA RESTRICTED No. IRS DIFFUSION RESTREINTE Title - Titre Corrosion Damage.to the Reactor Coolant Pump Closure Studs Due to Boric Acid Attack Country - Pays Date of Incident -~Date de l' incident May 16,1980 United States Type of Reactor - Type de rsacteur PWR Plant - Centrale Licensee - Ditenteur du permis d' exploitation Omaha Public Power District Fort Calhoun
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Unit N*
- Tranch n* i Combustion Engineering Power - Puissance First Commercial Operation gj73 MWe(net)
Date de mise en service pg Systems or Components Affected - Systhmes ou composants affect 6s Reactor Coolant Pump Closure Studs Flexatallic Gaskets 1
Initial Plant Condition - Etat initial de la tranche Startup testing Way in which Incident was Detected ?
Ccament l' incident a-t-il,its d6tects ?
Routine Startup Inspection Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou libsration de radioactivitd None l
Date of Receipt - Date de tsception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetses et
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Corrosion Damage to Reactor Coolant Pump Closure Studs due to Boric Acid Attack Event Description On the evening of May 15, 1980, plant operating personnel were performing a cold pressure test of the reactor coolant system prior to plant startup.
The primary system was pressurized to approximately 180 psig using one low pressure safety injection pump. A routine primary system leakage inspection by operations personnel revealed leakage coming from reactor coolant pump RC-3C.
At this time the low pressure safety injection pump was secured.
On the morning of May 16, 1980, the system was.again pressurized using the low pressure safety injection pump. An inspection by maintenance personnel revealed leakage from the shaft seal on pump RC-3C. This seal was subsequently tightened which reduced but did not eliminate the leakage.
Further investigation revealed leakage emanating from the gasketed surface between the pump casing and the pump cover. At this time, technical assistance was requested from the pump manufacturer, Bryon Jackson.
On the afternoon of May 16, 1980 the other three reactor coolant pumps RC-3A, RC-3B, and RC-3D were inspected.
Leakage was found to be coming from RC-3A and RC-38.
In addition, it was noted that corrosion damage had occurred to a number of the closure studs on pumps RC-3A and RC-38.
No stud damage was evident on RC-3D.
The corrosion had occurred on the exposed shank of the studs.
The pump cover and the pump casing for each reactor coolant pump are constructed of ASTM A-351, Grade CF8M stainless steel. Sealing between the pump cover and the pump casing is accomplished by means of two concentric 304 stainless steel Flexitallic gaskets. The cover is secured to the pump casing by 16 closure studs constructed of ASTM A-193, Grade B7 carbon steel, chrome plated in the thread area, and phosphate coated in the shank area. The studs..are approximately 3b" in diameter and approximately 29" in length. During assembly the studs are hydraulically tensioned to approximately 23,500 psi. The residual stress in the studs, when they are held by the nuts, is approximately 15,500 psi.
A leakoff line is included between the two Flexitallic gaskets; this line was originally blanked off with a pipe plug. To aid in determining the source of the leakage, the leakoff line for each pump was unplugged and the system was repressurized to approximately 180 psig.
Pump RC-3A showed slight leakage through the leakoff line.
Pumps RC-3B and RC-3C showed considerably more leakage. No leakage was noted for RC-3D.
The covers were removed from pumps RC-3A, RC-3B, and RC-3C. The Flexitallic gaskets were inspected and it was determined that deterioration of the gaskets was apparent in some cases. Where deterioration was apparent, it was in the form of loss of asbestos filler, not catastrophic failure. The gasketed surfaces for all three pumps were in good condition with no evidence of steam erosion.
Following removal, the 48 used studs were subjected to ultrasonic, magnetic particle, and visual examinations. A stud was considered to be acceptable for further service if the ASME Section XI acceptance criteria were met and if corrosion had not reduced the minimum diameter to less than 3.320" (the minimum accept-able diameter for new studs). The following is the final status on the studs removed from service.
a
. Number of Studs Reason for Rejection RC-3A RC-3B RC-3D Excessive corrosion 4
8 6
Other unacceptable indications 2
2 0
3 IO I
Cause of Event Leakage from the pumps was found to be caused by deterioration of the Flexitallic gaskets. Damage to the closure studs is believed to have been caused by boric acid attack.
Reason for Reporting to IRS This occurrence is reportable pursuant to Criteria 2.2, " Degradation of the Primary Coolant Pressure Boundary Main Steam Line, or Feedwater Line," and Criteria 4, "Significant Generic Problems."
Action Taken The affected pumps were reassembled using new studs as required and new Flexitallic gaskets. The nuts were visually inspected prior to reinstallation and one was replaced owing to excessive thread damage.
In order to preclude future stud damage and unidentified gasket failures, a number of preventive measures have been adopted.
First, the leakoff lines between the primary and secondary gaskets have been connected to pressure transmitters so that leakage past the primary gaskets will.be detected. Second, the accessible portions of the pump closure studs will be visually inspected at least once per fuel cycle.
(The accessible portions were, in all cases, the positio.ns susceptible to corrosion damage.) The original insulation for the pumps has been replaced by insulation which will allow better inspection of the studs. Third, further investigation will be conducted in an attempt to ascertain the precise mechanisms responsible for the stud damage.
In order to ensure that wetting of the old insulation did not create the possibility of chloride attack to the pump casing, samples of old insulation were analyzed for Cl, Na+, and SiO2 concentration.
In all instances, the analyses verified that the insulation met the acceptance criteria in Regulatory Guide 1.36.
In addition, a sample area of pump casing was liquid penetrant inspected to co'nfirm that no evidence of stress corrosion cracking had occurred.
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OECD NUCLEAR ENERGY T
A AGENCY
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38, bd.Suchet75016 Paris INCIDENT REPORTING SYSTEM Til. 524.96.93 Telex 630668 AEN/NEA RESTRICTED No.lRS DIFFUSION RESTREINTE Title - Titre Stud Failure in Reactor Coolant System Pressure Boundary - Bolted Closures Date de l' incident Country - Pays Date of Incident March 12, 1982 United States Type of Reactor - Type de r6acteur PWR D6tenteur du permis d' exploitation Plant - Centrale Licensee Maine Yankee Maine Yankee Atomic Power Company Manufactuer - Fabricant Unit N*
- Tranche n*
Combustion Enoineerina Power - Puissance First Commercial Operation 82S Mwe(net)
Date de mise en service December 1982 Systems or Components Affected - Systbmes ou composants affect 4s Steam Generator Manway Bolts Initial Plant Condition - Etat initial de la tranche j
Shutdown for maintenance outage J
i Way in which Incident was Detected ?
Comment 1* incident a-t-il. 6td d6tects ?
j Routine maintenance Radiation Exposure or Radioactivity Release Exposition aux rayonnements ou lib 6 ration de radioactiviti Date of Receipt - Date de r6ception Date of Distribution - Date de distribution Event description, possible causes, actions taken or planned and lessons learned (safety significance of incident) should be included in the following pages.
nesc-iar rn de l ' in c idcent, causes possibles, mesures prises ou projetses et
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t' Event Description During a maintenance outage a bolt problem was discovered.
During routine disassembly of the number 2 steam generator cold leg manway, six of the 20 lb" x 10" studs failed.
An additional five showed cracks when ultrasonically tested. The SA 540 grade B24 alloy steel studs had been exposed to boric acid as a result of primary leakage.
In order to stop this leak the studs were torqued to the maximum allowed value of 900 ft/lb on July 7, 1981.
Subsequently, three applications of Furmanite Sealing Compound were added (October 25, 1982,
November 7, 1981, and December 20,1981) td seal the leak.
~
~
Af ter the studs were replaced on the number 2 steam generator a reactor coolant system leak test was performed. No leakage was detected from the manway.
During the leak test a small leak was detected from the reactor vessel head seal.
The licensee removed the reactor vessel head and replaced the two head seal 0-rings.
The studs which were wetted were nondestructively tested and found to be acceptable.
The plant was restarted on March 30, 1982.
Cause of Event Cause is unknown at this time, however, some sort of corrosion is suspected due to exposure to boric acid.
Reason for Reporting to IRS This event is being reported pursuant to categories 3 and 4, "Significant Deficiencies in Design, Construction, Operation or Safety Evaluation" and "Significant Generic Problems," respectively.
Actions Taken Metallurgical analysis.of the failed studs has been initiated to determine the failure mode. All 20 studs on the affected manway will be replaced.
Ultrasonic testing of the primary manway studs on the other steam generators revealed no problem s.
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OECD NUCLEAR ENERGY AGENCY INCIDENT REPORTING SYSTEM 38, bd. Suchet 75016 Paris Tel 524.96.93 Tetax 630668 AEN/NEA 4
RESTRICTED DIFFUSION RESTREINTE No. IRS Titre Title Bolt Corrosion Date of Incident - Date de l' incident Country - Pays Various U.S.
Type of Reactor - Type de reacteur Various Plant - Centrale Licensee - D6tenteur du permis d' exploitation Many
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Unit N*
- Tranche n*
Power - Puissance First Commercial Operation -
MWe(net)
Date de mise en service Systhmes ou composants affectds Components Affected Systems or Bolts Initial Plant Condition - Etat initial de la tranche Wa y in which Incident was Detected ?
u Comment l' incident a - t - l'1. 6 t 4 d i t e c t 6 ?
Radiation Exposure or Racioactivity Release -
Exposition aux rayonnements ou lib 6 ration de radioactivitd Date of Receipt - Date de rdception Date of Distribution - Date de distribution description, possible causes, actions taken or planned and lessons learned Event (safety significance of incident) should be included in the following pages.
Description de l' incident, causes possibles, mesures prises ou projetees et A cnrerA1_doiveng FMeurer
IRS Report on Bolt Corrosion A number of events have been reported r'ecently which involve cases of bolt corrosion and failure in various plant systens. Some of these can be attributed to boric acid attack. The attached events are grouped together as one report because they are indicative of potential generic con cern.
Attached Reports:
1.
Stud Failure in Reactor Coolant System Pressure Boundary - Bolted Closures.
2.
Corrosion Damage to the Reactor Coolant Pump Closure Studs Due to Boric Acid Attack.
3.
Core Barrel Assembly Thermal Shield Bolts Broken.
4.
Crack Indications on Steam Generator Primary Manway Studs.
5.
Cerrosion of Studs on Valve in Spent Fuel Pool.
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