ML20070B863
| ML20070B863 | |
| Person / Time | |
|---|---|
| Issue date: | 11/30/1982 |
| From: | Massaro S NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-BR-0051, NUREG-BR-0051-V04-N4, NUREG-BR-51, NUREG-BR-51-V4-N4, NUDOCS 8212130140 | |
| Download: ML20070B863 (30) | |
Text
NUREC/BR-0051 f21 POWER REACTCiR EVENTS k!
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United States Nuclear Regulatory Commission May-June 1982/Vol. 4, No. 4 Power Reactor Events is a bi-monthly newsletter that compiles operating experience information about commercial nuclear power plants. This includes summaries of noteworthy events and listings and/or abstracts of UsNRC and other documents that discuss safety-related or possible generic issues. It is intended to feed back some of the lessons learned froen operational experience to the various plant personnel, i.e., managers, licensed reactor operators, training coor-dinators, and support personnel. Referenced documents are available from the USNRC Public Document Room at1717 H Street, Washington, DC 20555 for a copying fee. Subscriptions and additional or back issues of Power Reactor Events may be requested from the NRC/GPO Sales Program,(301) 492-9s30, or at PHIL-016, Washington, DC 20555.
Table of Contents Page SUMMARIES OF EVENTS 1.0 Overexposure During Entry into Reactor Cavity with incore instrumentation Thimbles Retracted..
1 2.0 Stuck Traversing incore Probe Causes High Local Radhtion--
7 3.0 Temporary Unavailability of Diesel Generator Emergency Power..
9 4.0 Inoperable Overpressure Protection System.-
10 6.0 Loss of Salt Water Cooling System.
12 6.0 Broken Reactor Coolant Pump Diffuser Bolts;..
16 7.0 Cracked Hydraulic Speed Control Cylinders on MSIVs......
16 8.0 References--
18 ABSTRACTS OF OTHER NRC OPERATING EXPERIENCE DOCUMENTS Abnormal Occurrence Reports (NU R EG-0000) Issued in May-June 1982... --
20 Bulletins, Circulars, and Information Notices issued in May June 1982....
22 Operating Reactor Event Memoranda Issued in May-June 1982--
27 l
Engineering Evaluations and Case Studies Issued in May-June 1982.
28 Editor: Sheryl A. Massaro Office for Analysis and Evaluation of Operational Data U. S. Nuclear Regulatory Commission Published in:
November 1982 Washington, D. C. 20555 8212130140 821130 PDR NUREG BR-OO51 R PDR
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NOTICE T0:
Plant Management and Training Personnel Corporate Contacts Other Recipients
SUBJECT:
PURPOSE AND DISTRIBUTION OF POWER REACTOR EVENTS (NUREG/BR-0051)
The bimonthly Power Reactor Events newsletter is an informal but important vehicle for the NRC to feed back the lessons of operational experience and to inform licensees and the nuclear industry of situations that may pose a safety concern to nuclear power plants. The document does not impose require-ments on licensees.
Because of the large volume of operational experience documentation produced each month, both foreign and domestic, and in view of the limited resources available at individual plants to assess and become familiar with items of interest, the format and content of Power Reactor Events have been revised to improve its usefulness and value to plant operating staff (i.e., management, training staffs, operations and maintenance personnel, and other technical and quality assurance staffs).
Starting with this issue (Vol. 4, No. 4), the following changes are being introduced: (1) the number of events selected for coverage has been increased, with emphasis continuing to be on events that have important lessons which may be useful to your facility; and (2) each issue will have a bimonthly " digest" of other documents involving operational experience produced by the NRC during the report period. This new section will list and abstract:
- Information Notices, Circulars, and Bulletins
- Operating Reactor Experience Memoranda and Generic Letters Case Studies, Engineering Evaluations, and Abnormal Occurrence Reports
- Pertinent documents from Regulatory and Technical Reports (NUREG-0304)
Power Reactor Events will continue to be provided free of charge to Plant Managers, Plant Training Officers, and Corporate Contacts for each nuclear power plant having an operating license or construction permit.
In addition, Federal agencies, certain State agencies, and certain foreign governments will receive the document free of charge. All other recipients have been notified in previous issues of the new subscription rates and should have submitted the order form provided to avoid interrupting receipt of the document.
Another NRC publication that may be of interest to you is the Licensee Event Report (LER) Compilation (NUREG/CR-2000). LERs are submitted to the NRC by nuclear power plant licensees in accordance with requirements in the license technical specifications. The monthly LER Compilatiun contains abstracts of those reports processed during the month into the LER data file of the Nuclear Operations Analysis Center (NOAC) at the Oak Ridge National Laboratory. The LER abstracts are arranged alphabetically by facility name, and then chrono-logically by event date for each facility. Component, system, and keyword indexes follow the summaries. The report is prepared by NOAC for the NRC's Office for Analysis and Evaluation of Operational Data. Eligibility for free distribution of the document is the same as for Power Reactor Events.
Inquiries regarding the subscription service for either document should be directed to the NRC Sales Manager at (301) 492-9530. Any questions regarding the content of the LER Compilation should be directed to Joel Buchanan, NOAC,-
at (615) 574-0391; questions regarding the content of Power Reactor Events should be directed to Sheryl Massaro,llRC/AE0D at (301) 492-4499.
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i; SUMMARIES OF EVENT 9 1.0 OVEREXPOSURE DdRING ENTRY INTO REACTOR CA ITY VITH INCORE INSTRUMENTATION THIMBLES RETRACTEI),
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On March 25, 1982, a ' shift engineer,at Zion Unit 1* received a radiation exposure of about 5 rem during an entry into the reactor cavity area with the incore instrumentation thimbles retracted.
This exposure was in excess of the permissible limit.s specified,in Title 10, Part 20 of the Code of Federal Regulations, i.e., 1.25 rem per calendar quarter.
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General Plant Conditions and Sequence of' Occurrences Unit 1 was in cold shutdown for refueling and maintenance.
Incore instrumenta-tion thimble retraction from the reactor core was started during the evening shift on March 23, 1982, and was completed about six hours later at approxi-mately 4:00 a.m. on March 24. A maintenance procedure, " Retracting and Insert-ing Incore Instrumentation Thimbles," requires that all access doors to the reactor cavity be locked with "R" locks, and all incore, detectors be in the storage position before the thimbles are retracted.
Control of keys to the "R" locks is administrative 1y assigned to' the shift-engWer on du,ty. ',
Shortly after thimble retraction was completed, the licensee began to fidod the refueling cavity in preparation for refueling.
At'abcut10:3/a.m.,'i4 was determined that the water level in the refueling cwJty was decreas'pi.#
Ataboutnoon,ashiftforemanbrieflyenteredthecav)tybeneaththereactor-vessel (hereaf ter called reactor cavity) in an effort to locate the leakag2) f source.
The shift foremac,4aw that the leakage was(significant.
The licensee decided to lower the water in the refueling ca,fity,At about 11:00 p.m. ther &tqtall the vessel head, and investigat/,the Teakage sourcs.
licensee found an excore nuclear instrumentation cover gasket,had slipped and was apparently the cause of the leak.
- s After the gasket was repiaced, the licensee raised the vessel head and flooded the refueling cavity to about 130 inches.
At about 6:00 p.m. on March 25, a shift engineer entered the reactor cavity to determine if there was further leakage.
During this entry, the shift engineer received a radiation dose'in L
excess of regulatory limits.
Tne leakage continued.
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The licensee raised the water 3tvel to see if increased static water head would seat the gasket and stop' the leak.
At about 9:30 p.m. a shift foreman briefly entered the reactor ca5ity. ud found there was still leakage.
The licensee again lowered the refueling catity water level.,
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- Zion Unit 1 is a 1040 M@ PWR located 40 mileEiiorth' of Chicago, Illinois, and is operated by Commonwealth Edison. 4
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I After further gasket replacement on March 26, the refueling cavity water level g
was again raised.
At atout 6:00 a.m. on March 27, it was determined that there was still leakage.
The licensee again lowered the refueling cavity water level.
After installing redesigned gaskets on the nozzle flanges, the licensee inserted
,the incore lastrumentation thimbles during the shift day on March 28, and again raised water / level in the refueling cavity. With the thimbles inserted and radiation "evels in the reactor cavity greatly reduced, entry was made under the reactor / vessel to check for leaks.
No significant leakage was identified.
The licensee retracted the incore instrumentation thimbles and proceeded witn the refueling.
[ReactorCavityEntries On March 24, a shift foreman obtained an administrative approval for dose exten-sion (to 500 mrem for the day) frcm a plant health physicist, and a digital dosi-meter from a rad / chem foreman.
He then proceeded to the reactor cavity access area where a' rad / chem technician (RCT) trainee was already monitoring the installation of a temporary pump in the cavity.
According to the licensee, entry into the cavity was not made while installing the pump; the pump was lowered by rope.
The RCT trainee stated that he had been informed by the rad / chem foreman that a cavity entry would be made.
The rad / chem foreman cautioned the trainee to be careful because high radiation levels may be encountered in the reactor cavity.
The trainee did not make a e
radiological survey in the reactor cavity before the shift foreman arrived, however, and there was no discussion between the RCT trainee and the shift fore-man concerning radiological conditions in the reactor cavity before the foreman made the reactor cavity entry.
The shift foreman borrowed the RCT trainee's R0-2 portable survey instrument and made an entry into the cavity down to the bottom of the ladder.
The shift foreman had the R0-2 on its lowest scale (0-500 mR/hr) during the descent, and said that he did not look at the R0-2 meter on the way down.
As he neared the bottom'of the ladder, he was alerted to increasing radiation levels by t 5 audible indication of the digital dosimeter, and he glanced at the survey meter as he' reached the bottom of the ladder.
Upon seeing that the meter was off scale, he immediately climbed out of the cavity.
The shift foreman's dosimeters indicated an estimated exposure of about 150 mrem during the entry.
Afterdheshiftforemanmadetheentry,theRCTtraineewentdowntheladderto about Point B (see Figure 1), where his'R0-2 meter pegged full scale on the 0-5
,\\(R/hrscale.
The RCT trainee made another entry with a teletector.
He went down the ladder to about Point A, extended the teletector probe, and read exposure a
rates of 35 R/hr at Point B and 85 R/hr at Point D.
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, entries were made on March 24.
Another entry was made on March 25 to replace the excore cover gasket, and plans were made to again increase refueling cavity water level.
An operations engineer wrote a night order which stated, "With water above the flange, make
,'an entry to the cavity area with RP [ radiation protection] and check for leaks as best as possible minimizing exposure."
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At about 6:00 p.m., the shift engineer informed the rad / chem foreman that he was preparing to enter the reactor cavity, and also requested approval from the plant health physicist for dose extension to 500 mrem for the day.
The shift engineer was wearing a 0-200 mR self-reading dosimeter and a film badge.
There was no discussion concerning the need for additional dosimetry.
The rad / chem foreman assigned an RCT to cover the job.
The foreman and the RCT recalled discussions about an exposure rate of 85 R/hr from the previous day's entry, conducted on another shift, but they were unable to find the survey record to verify this information.
The foreman later assigned an RCT trainee to assist the RCT.
(The rad / chem foreman, the RCT, and the RCT trainee assigned lacked experience in radiological monitoring and control in high radiation areas, and were unfamiliar with the reactor cavity area.
Other RCT and plant personnel on duty had significantly more experience.) When the shift engineer went past the rad / chem foreman while leaving the office, the foreman asked if the incore detectors were " parked." The shift engineer responded "yes."
There was no discussion of thimble position.
The involved health physicist and rad / chem foreman stated that each had assumed that the other had discussed radiological planning for the entry with the shift engineer.
They both stated that the shift engineer was more familiar with the area than they were.1 In preparation for the entry, the RCT attempted to locate a teletector that was calibrated on its top scale.
The RCT was not successful in locating such a teletector and went to the reactor cavity access area with a teletector and an R0-2A survey meter that were calibrated to 50 R/hr.
The RCT and shift engineer were wearing full protective outer clothing with plastic rain suits and full face respirators.
The shift engineer was also wearing rubber boots because he expected that there would be water above the cavity platform (Platform #1 in Figure 1).
The RCT took the teletector and a flashlight and proceeded down the esvity ladder (Ladder #1 in Figure 1) to make a survey. When he reached Point A (Figure 1), he read an exposure rate of about 200 mR/hr.
He said that this surprised him because he was expecting 85 mR/hr.
.de RCT then extended the teletector probe down and in front of the ladder and read an exposure rate of about 35 R/hr at Point B and about 50 R/hr at Point C.
The RCT then handed the teletector up to the RCT trainee, who was above at the top of the ladder, took the R0-2A from the trainee, and went down the ladder to the bottom step.
The RCT extended his arm to about Point C (three feet above the platform), and verified the 50 R/hr reading.
No further surveys were taken.
The RCT stated that he then went up the ladder to about Point A, called the exposure rate at Point C to the RCT trainee, and told the shift engineer he could now go down.
When the shift engineer arrived at about Point C, the RCT called to the trainee to start keeping time.
The shift ergineer was told the dose rate at Point C (50 R/hr) but was not told his allowed stay time, nor was there any discussion of his intended actions in the reactor cavity.
The RCT trainee calculated the permitted stay time to be 30 seconds (about 400 The mrem) to keep the shift engineer within his dose extension of 500 mrem.
4
shift engineer descended the ladder to the platform which was covered with about six inches of water.
The shift engineer then waded in toward the bottom of the reactor vessel.
The shift engineer estimates he went at most eight feet along the platform.
When the trainee yelled that 30 seconds was up, the RCT called to the shift engineer to come out.
When the shift engineer failed to show up in a few sec-onds, the RCT called again and went further down the ladder.
The RCT saw the shift engineer wading back toward the ladder.
The RCT and shift engineer then climbed out of the cavity.
The trainee stopped the stopwatch at 67 seconds when he could see the shift engineer on the ladder.
The RCT returned to the rad / chem office and told the lead health physicist and the rad / chem foreman that the shift engineer received an estimated dose of 900 mrem, based on 67 seconds in a 50 R/hr field.
The lead health physicist took the shift engineer's film badge and told him not to enter the controlled area until the dose had been evaluated.
The film was sent to the vendor on March 26.
The RCT said that he did not expect the exposure rate to increase as the shift engineer approached the reactor vessel.
The RCT was not knowledgeable about the source of radiation in the reactor cavity or the anticipated radiation levels.
The shift engineer said that he was aware that the exposure rate would increase as he approached the bottom of the reactor vessel.
He also said that when he decided to leave the ladder and walk toward the bottom of the reactor vessel to look for the source of leakage, it was difficult to hurry because the water was about six inches above the platform and his rubber shoe covers were only eight inches high.
After the shift engineer's entry into the reactor cavity, the licensee raised the water level in the refueling cavity to see if increased static head would seat the gaskets.
At about 9:30 p.m.,
the same RCT and RCT trainee monitored for a cavity entry to be made by the shift foreman.
The shift foreman made a brief entry to about Point B and saw that there was still significant leakage.
Personnel Overexposure On March 27, the film badge vendor reported that the shift engineer's film badge reading was 3700 mrem.
Because of the configuration of the reactor cavity and the location of the active portion of the withdrawn thimbles, it was probable that the dose received by the individual's lower trunk was greater than to the film badge.
The licensee subsequently estimated a dose of about 4.7 rem for the entry.
An independent evaluation by NRC inspectors indicated a similar calculated dose of about 5 rem.
Inspection Results The results of an NRC inspection conducted on March 30-31, April 7-8, and April 29 indicated serious weaknesses in the licensee's radiation protection 5
program concerning systematic evaluation and planning of radiation work.
Specific weaknesses included:
(1) lack of coordination between plant health physicists and rad / chem foremen in planning the entries, (2) inadequate radia-tion surveys associated with the entries (the_ vertical and horizontal profiles of dose rate in the cavity would have been useful later on), (3) use of inexpe-rienced rad / chem technicians to monitor the entries, (4) lack of understanding by radiation protection personnel of the reactor cavity radiological hazards including the radiation sources, (5) inadequate training in reactor cavity radiological hazards even though a similar overexposure had occurred in 1976, (6) failure of shift operations personnel in leadership positions to exhibit good radiation protection practices, and (7) unavailability of survey instru-ments calibrated to greater than 50 R/hr.
The Zion licensee was fined $100,000 because of what the NRC believed to be a serious breakdown in management controls and the radiation protection program, resulting in an unnecessary exposure.
The NRC was particularly concerned that this overexposure occurred even though the licensee had implemented actions to avoid such an occurrence in response to the NRC's IE Circular No. 76-03, "Radia-tion Exposures in Reactor Cavities," September 13, 1976, based on a similar over-exposure at Zion Unit 1 in 1976.
(In addition, a similar event occurring at Salem Unit 1* was summarized in Power Reactor Events, Vol. 3, No. 2, " Personnel Exposures During Entry into Thimble Area," pp. 10-11.)
The licensee's corrective actions include:
(1) requiring that thimbles be reinserted into the reactor core prior to personnel entry into the reactor cavity beyond the base of ladders extending into the cavity area, that a special lock be placed on the door to the cavity when the thimbles are removed, and that locations of thimbles and incore detectors during outages be posted in the rad /
chem office; (2) reviewing and supplementing operations and radiation personnel training;** (3) instructions to personnel on the importance of good communica-tions requiring a radiation work permit (which includes a written description of work to be performed) for individual jobs exceeding,50 mrem; (4) revising radiation protection procedures to include specific requirements for issuing high range (} 500 mrem /hr dosimeters; and (5) maintaining in the rad / chem department addition high range detectors calibrated to 1000 R/hr, and a lirited number of dose rate ionization chamber instruments with lighted dials for >"rk in dark areas.
These actions are being taken for both Units 1 and 2.1 4
- Salem Unit 1 is a 1090 MWe PWR located in New Jersey, 20 miles south of Wilmington, Delaware, and is operated by Public Service Electric and Gas.
- After the March 17, 1976 overexposure, the licensee instructed station per-sonnel about the incident, the cause and the radiological hazards.
- However, specific instruction was not included in ongoing training on hazards in the reactor cavity. Training was general in nature with no specific description of radiation sources or expected rapid exposure rate changes when the thimbles are withdrawn.
6
a 2.0 STUCK TRAVERSING INCORE PROBE CAUSES HIGH LOCAL RADIATION On June 3,1982, with reactor power level at 75%, a traversing incore probe (TIP) at Pilgrim Unit 1* retracted beyond its normal shielded storage chamber, causing high radiation levels within the drive mechanism area of the reactor building.
These levels exceeded the limits specified in the plant's emergency plan.
A site alert was declared per Pilgrim's emergency procedures and subsequently terminated when radiation levels were verified to be localized and decreasing.
Due to a faulty cable connection, th'e area radiation monitor (ARM) for the TIP machine area did not alarm when it should have. The over-retraction of the probe was ideatified by a technician in the control room using a remote TIP position indicator.
The watch engineer and health physicist were notif#,ad immediately and a radiation survey was conducted.
The area was declared a high radiation area (50-90 R/hr on contact with the TIP drive mechanism, 15 R/hr at 10 ft, and 350 mR/hr at 20 ft) and was guarded and barricaded.
Surveys indicated that the TIP detector was on the drive wheel within the TIP drive housing.
A portable shield was placed to permit access to the reactor core isolation cool-ing (RCIC) system area.
Repairs were delayed until the source decayed to levels justified by the station's ALARA policy.
No overexposures to personnel or abnor-mal releases occurred.
By 8:30 a.m. of the following day, the dose rate on con-tact with the TIP drive mechanism had decreased to 1.8 R/hr and continued to decrease.
The probe itself is attached to a flexible drive cable which is driven from outside the primary containment (see Figure 2).
Through the use of an indexing mechanism, the cable is routed into any one of ten guide tubes which continue into the reactor core.
A position limit switch provides an electrical inter-lock release when the probe is in the nominal zero position to allow the index-ing mechanism to index the TIP to the next guide tube location.
TIP overtravel was attributed to failure of a faulty limit switch and guide tube misalignment.
Af ter radiation in the drive mechanism area decreased to a sufficient level, repairs to the limit switch and to the guide tube slot alignment were made.
The faulty cable connection for the ARM was also repaired, and the TIP was returned to operation.
The licensee revised operating procedures to improve response to this type of occurrence.
Review of the event indicated that ARM calibration procedures did not require a functional check with a source after calibration and reinstallation of the detector.
The licensee is revising his procedure to include a functional check, and is also emphasizing the existing precautions included in the TIP calibration and operating procedures to preclude withdrawing the TIP detector past its chamber shield.
Following termination of the site alert, the licensee kept a health physics technician at the entrance to the T7.P machine area because of an inability to lock the area.
The NRC inspector toured the area at 6:50 a.m. on June 4, 1982 and found the technician asleep.
Apparently, the technician had an injured back and had taken medication which contributed to drowsiness.
The licensee is reviewing his medical department's policy and practices and has discussed the
- Pilgrim Unit 1 is a 670 MWe BWR located four miles southeast of Plymouth, Massachusetts, and is operated by Boston Edison.
7
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g incident with his Health Physics Staff The NRC inspector cited a violation involving failure to lock or control access to a high radiation area.E 7 3.0 TEMPORARY UNAVAILABILITY OF DIESEL GENERATOR EMERGENCY POWER An incident occurred on June 2, 1982 which resulted in the loss of cil three diesel generators at Calvert Cliffs.* With Calvert Cliffs Unit 1 in a refuel-ing outage, and Unit 2 operating at full power, diesel cenerator no. 11 (Unit 1) was removed from service for maintenance.
At 5:30 a.m., in preparation for per-forming maintenance on a service transformer, the remaining diesel generators, no. 12 (shared) and no. 21 (Unit 2), were tested and shown to be operable.
Following removal of the service transformer, dicsel generator no. 21 was paral-leled to offsite power and fully loaded.
At 6:32 a.m., diesel generator no. 21 tripped and diesel generator no. 12 was started.
At 7:05 a.m. diesel generator no. 12 tripped; thus, all diesel generators were temporarily inoperable.
Diesel generators nos.12 and 21 were returned to operability at 7:35 a.m and 8:00 a.n;.
respectively.
At 2:45 p.m. diesel generator no. 21 tripped again on loss of field during a load reduction, af ter a :;uccessful one-hour surveillance test run.
It was started and declared operable at 3:15 p.m.
Offsite power remained in service during the event.
Diesel generator no. 21 tripped due to voltage regulator drift while operating parallel to offsite power.
The voltage regulator drift produced a leading reactive load resulting in a sensed loss of field, tripping its protective relay.
Diesel generator no. 12 tripped when an operator raised the voltage of the Unit 2 generator that it was in parallel operation with, which also produced a leading reactive load and resulted in a sensed loss of field, tripping its protective relay.
Diesel generator no. 21 voltage regulation had normally exhibited greater tend-ency to drift when operating in parallel with an offsite power source than h;4 that of nos. 11 and 12. When acting as the single supply to a vital bus - tk design condition of the diesel generators during an accident - each diesel generator exhibits adequate regulating capability.
Several portions of the voltage regulator circuit, which allow parallel operation, are not active in the circuit logic during design conditions.
In addition, the loss of field trip is not active when the diesel generator is not paralleled to other sr..ct-and would not have resulted in a trip during design conditions.
1 Control room instrumentation does not include either power factor or leadi g l
reactive load indication (available locally), since these indications would not be necessary during design conditions.
However, when the diesel gens,tcrs are run parallel with other generators, monitoring of such load condit w becomes necessary.
Operations personaal reported that a caution note, d..eck ing that the operator be careful of diesel generator loads when changing udn generator voltage, had been r? moved prbr to this event in an effort to c' inn up the control boards.
Also, an NRC inspactor noted that the licensee's
- Calvert Cliffs Units 1 and 2 are 825 MWe PWRs located 40 miles south of Annapolis, Maryland, and are operated by Baltimore Gas and Electric.
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Operating Instructions do not address the diesel generator trip on a leading power factor and do not adequately address actions to avoid such a trip.
The voltage regulator assembly for diesel generator no. 21 will be replaced to improve regulation. A caution statement will be added to the operating instruc-tions, and a permanent caution label will be affixed to the control room control panel alerting operators to monitor diesel generator reactive load when adjust-ing system line voltage.
This permanent label will also serve to remind opera-tors of the necessity to monitor reactor lond during parallel operation of diesel generators.8-to 4.0 INOPERABLE OVERPRESSURE PROTECTION SYSTEM Portions of the reactor nyerpressure protection system (OPS) were inoperable at North Anna Unit 1* from May 19 until May 22, 1982.
The reactor sas in cold shutdown during this period, and the OPS was not called upon to activate.
The reactor coolant system (RCS), which the OPS is designed to protect from over-pressurization during low temperature conditions, was in a solid water condition on May 21 and 22.
On May 19, the nitrogen pressure in the A OPS reservoir decreased to below allow-able limits to operate PORV 1456.
This rendered the PORV inoperable for use as overpressure protection.
(One inoperable PORV is permitted for seven days by the plant's technical specifications.) However, on May 22, the low pressure alarm activated for the B nitrogen reservoir, which sLpplies the remaining PORV 1455c.
A containment entry was made to manually cepressurize the B nitro-gen reservoir.
Another low pressure alarm occurred on the B OPS reservoir approximately an hour later, and a second containment entry was made.
It was reported that the B nitrogen reservoir was isolated from the system due to an isolation valve being closed, thus rendering the second PORV inoperable.
The nitrogen system is designed as a backup system for operation of the PORVs on the loss of instrument air during power operation (see Figure 3).
During cold shutdcwn conditions, the OPS is placed in operation by placing the PORV key switch in " auto." With the key switch in this position, the nitrogen system becomes the only motive force for opening the PORVs automatically on an over-pressure condition.
The operator who made the first containment entry on May 22 stated that he found the B reservoir isolation valve closed.
The licensee stated that the last opportunity for manipulating that valve was on May 14. The licensee promptly reported that both PORVs were inoperable from May 19 to May 22, but believes it possible that the valve was mistakenly closed by the operator during the May 22 containment entry.
- North Anna Units 1 and 2 are 890 MWe PWRs located 40 miles northwest of Richmond, Virginia, and are operated by Virginia Electric and Power.
10
i TO PORV 1466 TO PORV 1455C J L J L
\\
"B" TANK l
l SOLATION VALVE f
f i
k U
D
'I*
"A" TANK ISOLATION VALVE g
(
r) l 4
y 80 a
l gg x
FROM N HEADER *_
"A" TAN K "B" TAN K FILL VALVE FILL VALVE "B"
"A" l
N TANK N TANK 2
2
/
\\
/
\\
l Figure 3 NITROGEN SUPPLY FOR NORTH ANNA, UNIT 1 i
l,
The total period of time for a single inoperable PORV was approximately 6 days and 11 hours1.273148e-4 days <br />0.00306 hours <br />1.818783e-5 weeks <br />4.1855e-6 months <br />.
An engineering evaluation performed by the licensee indicated the time involved for the two inoperable PORVs was 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and 45 minutes.
Either PORV could have been opened manually if required by activating the instrument air supply to the PORVs.
The licensee has developed valve lineup instructions, and control room drawings have been annotated. A training handout also has been developed as required reading for operators.
The licensee found the piping arrangement to be slightly different on North Anna Unit 2, and the drawings were incorrect.
The related I
l problems on Unit 2 were corrected the same time as were those for Unit 1.11 13 5.0 LOSS OF SALT WATER COOLING SYSTEM At San Onofre Unit 1* on May 13, 1982, during removal of the south salt water cooling pump (SWCP) for routine maintenance, the salt water intake structure flooded to a level six inches below the SWCP motor pump flange.
Sufficient heat removal capability was provided with both north and south screen wash pumps and the auxiliary SWCP to preclude heatup of the reactor coolant system.
The reactor had been shut down since February 27, 1982, for various inspections, modifications, and maintenance.
The reactor coolant piping had been drained for steam generator tube inspection, and the primary manways had been removed.
The upper component cooling water heat exchanqer was removed from service and 9 pen, and the south SWCP motor had been removed earlier in the day.
The auxil-iary SWCP circuit breaker was removed for maintenance and the pump's flow path was isolated and cleared by permission slip.
The north SWCP was operating and removing reactor decay heat from the component cooling water system via the lower component cooling water heat exchanger. The north and south screen wash pumps were operable in their normal alignment.
(The screen wash pumps can be manually cross-connected in p ovide backup to the salt water cooling pumps, but they have a lower capacity and are not qualified as safety related equipment.)
Reactor coolant temperature to the residual heat removal heat exchanger was 124 F, component cooling water out of the component cooling water her.t exchanger was 62 F, and, although the salt water temperature was not recorded, it had been 59-62*F in this period.
Both SWCPs and both screen wash pumps are mounted vertically on pedestals in the ocean water intake structure.
The pedestals are approximately three feet i
high, except for the north screen wash pump, which is approximately four feet high. The auxiliary SWCP is located in a separate enclosure outside of the f
intake structure.
At approximately 8:00 a.m., after removing the south SWCP motor and the nuts attaching the pump to its pedestal, a crane operator removed the pump.
Ocean
- San Onofre Unit 1 is a 436 MWe PWR located five miles south of San Clemente, California, and is operated by Southern California Edison and San Diego Gas and Electric.
12
water immediately entered the intake structure through the resulting hole in sufficient quantity to prevent personnel from reseating the pump on its f
foundation.
Flooding continued until 8:42 a.m. when the level in the intake structure rose to sea level, approximately five feet above the floor of the intake structure.
At this time maintenance personnel were able to partially remount the south SWCP and begin to slowly reduce the water level in the intake structure using a portable pump.
All salt water cooling was lost from 8:18 a.m. until 8:42 a.m., at which time the operators completed manual valve alignment to allow the north screen wash pump to supply salt water cooling to the lower component cooling water heat exchanger.
A discharge pressure switch which causes the north screen wash pump to automatically start was submerged, but this apparently did not affect opera-tion of the pump.
At 8:29 a.m. a spare breaker for the auxiliary SWCP was installed and the auxiliary SWCP was considered available if needed.
The isola-tion of the flow path of this pump by clearance slip was not removed until 6:25 p.m.; thus, the pump was not administratively operable durina this period, but could have been made operable by modifying the clearance slip.
Licensee per-sonnel stated that they did not do this because the north screen wash pump was operating and adequately removing decay heat from the reactor.
Salt water flow was lost from the north SWCP because the control operator secured the north SWCP when the pump amperage and the pump discharge valve began to cycle erratically.
When flooding ceased, the intake structure water level was approximately two inches below the pump motor vents for the north salt water and the south screen wash pumps, and approximately 18 inches below the north screen wash pump motor vents.
During the 24-minute period when ne salt water cooling was available, reactor coolant core outlet temperature did not rise perceptibly, reactor coolant inlet temperature rose 1.5 F, and component cooling water outlet temperature rose 15 F.
The north SWCP was returned to service at 1:14 p.m., but failed at 2:34 p.m.
because of a failed-shut discharge valve.
Salt water cooling was briefly inter-rupted, until the north screen wash pump was aligned to supply salt water cool-ing.
This abnormal alignment was maintained until 6:45 p.m. when the north SWCP was returned to service.
Apparently, the pump discharge valve failed closed due to residual moisture in the pressure switch and melted insulation in an isolated time delay relay.
No technical specification requirement exists which requires y
the salt water cooling system to remain operable while the reactcr is in cold
- shutdown, f
Equipment control forms were prepared by the licensee to plan refueling interval maintenance on the south salt water cooling pump and motor.
These forms indi-cated that a meeting was held on May 12, 1982 between maintenance and operations personnel to discuss the removal of this equipment from service.
No special precautions to prevent flooding upon pump removal were indicated at that time.
These forms referenced a permission slip, which indicated the valve alignment required by the watch engineer to perform this work.
This permission slip did not include isolating the sea water inlet to the intake structure.
13
i 4
The procedure specified for the maintenance work had one precaution to prevent flooding:
"To prevent flooding, remove and install the pump at the low tide only." This precaution was inadequate to prevent flooding because no measurable 4
tide level was specified.
4 Maintenance personnel had used a tide chart for Los Angeles (Outer Harbor),
45 miles northwest of San Onofre, to estimate the time of low tide as 8:06 a.m.
on May 13, 1982.
The tide level on this chart was indicated to be 0.3 feet mean low low water (MLLW). The maintenance foreman and watch engineer discussed this 1
level and estimated that it was approximately two feet up the three-foot tall i
pedestals of the salt water ~ cooling pumps.
The water level reached, however, was three feet higher than this.
These errors resulted in the work starting, and were a principal cause for the event.
t The control operator authorized removal of the south SWCP motor on the evening of May 12, under the permission slip.
This authorization was documented on the permission record sheet as required by the station procedure, " Equipment Control Implementation." The maintenance foreman directly supervising the pump removal, and the personnel removing the pump, beliend that this authorization included authority to remove the pump as well.
TV ; belief was based on their discussions with the previous shift watch engine
, the meetings held to discuss the work, and the permission holder tag for the permission slip which stated that it was for the south SWCP.
These personnel also noted that there were no other flooding precautions than the one mentioned concerning low tide in the approved procedure for the work.
However, the control operator had not docu-mented approval of the work as required by procedure, and had, presumably, not given approval.
Nevertheless, the work was begun.
Technical Specification 6.8.1, and ANSI Standard N 18.7-1976, Paragraphs 5.3.5(1) and 5.2.6 require that one consider the potential for flooding of the pump area of the ocean water intake structure.
The licensee's maintenance pro-cedure did not adequately consider the flooding potential of the work.
As a direct consequence, the salt water cooling system was effectively flooded and decay heat was not removed from the component cooling water system for 24 min-I utes.
In addition, the specific system designed to remove decay heat from the reactor (the salt water pumps, heat exchangers, and piping) was inoperable for two periods totaling about nine hours.
The NRC and licensee investigations of this event identified the following primary causes:
inadequate communication between the operators and maintenance personnel, an error in estimating the low tide level, and a ;;:aintenance proce-dure which does not fully address a necessary precaution.
Although a work authorization letter had been issued for scheduling the maintenance on the SWCPs, the maintenance workers and operators had failed to adequately review the hazards involved with respect to potential flooding before allowing the pump to be pulled.
Additionally, it has been common practice to pull an SWCP only with a circulating water pump running to depress the intake water level and help prevent flooding.
This precaution was not adequately called out in plant procedures, and was not employed by the maintenance personnel.
The licensee is upgrading the applicable procedures and has requalified main-tenance and operations personnel in the proper implementation of equipment control and the work authorization process, stressing that compliance is 14
1 essential to prevent incidents of this type.
Formal checklists for equipment removal and restoration will be established for the salt water cooling system.
On June 16, 1982, the NRC Region V office cited the licensee with a notice of violation for the failure to provide sufficient detail in the SWC maintenance procedure to safely account for the effect of ocean tide conditions.
The fact that the SWCPs were inoperable was not a violation since the license technical specifications contained no provision which requires the salt water cooling system to remain operable while the reactor is in a cold shutdown condition.
Subsequently, on August 13 and 19,1982, two more incidents occurred involving this system.
On August 13, while restoring the south SWCP to service (from the May 13. event), with the north SWCP supplying flow to the lower component cooling water heat exchanger, the discharge valve on the south SWCP opened unexpectedly.
l This allowed much of the flow from the running north SWCP to move in reverse
,~
through the idle south SWCP, bypassing the heat exchanger.
This valve was immediately shut by an operator at the valve, restoring full flow to the heat
)
exchanger.
The brief reduction of the flow through the heat exchanger did not observably increase core temperatures.
However, the incident was of concern because the pump discharge valves had been observed to be occasionally unreli-l able and erratic.
This was first noted by the NRC during the March 10, 1980 l
event review, and more recently following reviews of January 18, February 1 and j
March 19, 1982 incidents.
On August 19, 1982, the north SWCP had to be removed from service due to a smoking lower motor bearing. At the time of this incident, the south SWCP was out of service from the May 13, 1982 incident.
The nonsafety-related auxiliary SWCP was started to maintain sufficient salt water flow to the upper heat exchanger.
(This pump is hooked up to an emergency bus, but does not meet all seismic qualifications.) Temporary repairs to the north SWCP were made using a spare motor for this pump.
During this period of several hours, the nonsafety-related auxiliary SWCP and screen wash pumps were used to provide salt water flow.
Subsequent investigation of the north SWCP motor determined that the inside of the motor was rusty, muddy and oily, the lower motor bearing was wiped, and the pump upper bearing was excessively worn.
The losses and reductions of salt water cooling had no adverse effects on public health or safety.
However, as shown in a case study reportis prepared by the NRC's Office for Analysis and Evaluation of Operational Data for the avent on March 10, 1980, a complete loss of the salt water cooling system dur-ing the early stages of residual heat removal operation could lead to damage of some safety-related equipment within a few minutes.
Such single failure vulnerability of cooling water systems is under review as part of the NRC's Systematic Evaluation Program.
In addition, the NRC's Region V office is closely monitoring the licensee's engineering and human factors studies to improve the reliability of this system.
In the interim, additional operator training on special operation precautions for this system is planned.
15
6.0 BROKEN REACTOR COOLANT PUMP DIFFUSER BOLTS On April 23, 1982, the licensee at Robinson Unit 2* reported that four of 16 stainless steel bolt heads holding the diffuser casing to the adapter broke off when one of the reactor's three main reactor coolant pumps was being disassem-bled from the reactor coolant system for the ten year inservice inspection.
The licensee also identified apparent cracking on eight additional bolts.
On May 12, five of 16 stainless steel bolts were found to have missing hex heads when a second reactor coolant pump was disassembled, and eight bolts broke during its removal.
In both cases, the bolts were 5/8-inch in diameter by four inches in length, and were internal to the pump.
The licensee's investigation of the April 23 event revealed that the diffuser e
adapter bolts failed because of stress corrosion cracking.
High chloride levels were present, probably due to an externally introduced chloride contaminant and not system chemistry.
An investigation of the installation history for the bolts has shown that the diffuser adapter was removed and reinstalled in the field dur-ing plant construction.
It is therefore possible that the reinstallation of the' diffuser adapter in the field was improperly done, and that a chloride contami-nant was introduced at that time.
In an effort to determine the extent of stress corrosion, the casing adapter to casing bolts were examined.
A hardened coating of graphite lubricant on the bolts restricted the inspections to visual and ultrasonic examinations.
The visual inspection showed the bolts to be in good condition, and the ultrasonic examination showed no indication of cracking in the bolts.
The licensee there-fore concluded that the stress corrosion attack was limited to the diffuser adapter bolts, and replaced these bolts with new ones.
Most of the bolts remained intact during the April 23 and May 12 inspections, and thus the diffuser adapter could not separate and interfere with pump opera-tion.
This type of bolt failure increases the potential of a locked rotor transient.
The bolt failure remains under licensee and NRC review.is 21 7.0 CRAEKED HYDRAULIC SPEED CONTROL CYLINDERS ON MSIVs On June 2,1982, during normal operation surveillance testing on the main steam isolation valve (MSIV) closure instrumentation at Duane Arnold,** position switch 254415 did not provide a closure signal to the reactor protection system within the 10% valve closure limit.
As required by a technical specification, the MSIVs were closed within eight hours.
During the subsequent inspection, MSIVs CV4412 and CV4415 were found to have cracked hydraulic speed control cylinders.*** On CV4415, the cap screws holding the speed control valve manifold assembly to the j
hydraulic cylinder were broken and the assembly had fallen into the valve operator.
- Robinson Unit 2 is a 665 MWe PWR located five miles northwest of Hartsville, South Carolina, and is operated by Carolina Power and Light.
- These MSIVs are manufactured by Rockwell; the speed control cylinders are by Sheffer.
16
The licensee determined that the cracked MSIV hydraulic cylinders and failed speed control assembly cap screws were due to a violation of procedures during a May 22, 1982 startup.
The operator had failed to follow the operating pro-cedures for the opening sequence of the MSIVs; i.e., the four inboard MSIVs were opened before the four outboard MSIVs.
This action resulted in a pressure differential across the main disk while the outboard MSIVs were stroked open.
i The resulting force imbalance was absorbed by the hydraulic speed control unit which caused the damage described above.
The position switch was operable but out of adjustment, probably because of the pressure transient.
Before startup, the licensee plans to test and reset position switch ZS4415.
The four inboard MSIVs (CV4412, CV4415, CV4418, and CV4420) will be inspected and leak rate tested.
The cracked hydraulic cylinders and damaged speed control assembly will be repaired or replaced.
The four inboard MSIV actuators will be.
rebuilt.
A stress analysis performed by the licensee in'dicated that the piping between the inboard and outboard MSIVs was not overstressed; another analysis using the highest postulated forces will be performed to determine if any por-tion of the valve assembly was overstressed.
To prevent recurrence, plant procedures and operating instructions will be reviewed and revised as necessary, and a program will be established for additional operator training.22-24 17
l
8.0 REFERENCES
1 1.
Commonwealth Edison, Docket No. 50-295, Licensee Event Report 82-14, April 23, 1982.
2.
Letter from J. Keppler, USNRC/ Region III to J. O'Connor, Commonwealth Edison, transmitting Notice of Violation and proposed Imposition of Civil Penalties, July 9, 1982.
3.
USNRC/ Region III, Inspection Report 50-295, 82-09 (DETP), approved July 9, 1982.
4.
Letter from B. Lee, Jr., Commonwealth Edison to R. C. DeYoung, USNRC/IE responding to the July 9, 1982 letter (Reference 2), August 9, 1982.
5.
USNRC, Preliminary Notification PN0-I-82-40 (June 3,1982) and -40A (June 4, 1982).
6.
Letter from R. D. Machon, Boston Edison Company, to Director, USNRC/
Region I, June 4, 1982.
7.
USNRC/ Region I, Inspection Report 50-293/82-16, June 30, 1982.
8.
Baltimore Gas and Electric, Reportable Occurrences 82-25 and 82-27, June 2, 1982.
9.
USNRC, IE Inspection Reports 50-317/82-12 and 50-318/82-15, June 17, 1982.
10.
Baltimore Gas and Electric, Licensee Event Report 82-027/03L, June 21, 1982.
11.
USNRC memorandum from J. P. O'Reilly, Region II, to J. L. Crooks, AE00, transmitting proposed Enclosure 3 Item to Abnormal Occurrence Report, July 22, 1982.
12.
Virginia Electric and Power Company, Docket No. 50-338, Licensee Event Report 82-41, July 19,1982.
13.
USNRC, IE Information Notice 82-17, "0verpressurization of Reactor Coolant System," June 11, 1982.
14.
USNRC, Preliminary Notification PNO-V-82-025, May 13, 1982.
15.
Southern California Edison, Docket No. 50-206, Licensee Event Report 82-015, June 14, 1982.
16.
USNRC/ Region V, Inspection Report 50-206/82-17, June 15, 1982.
17.
NUREG-0090, Vol. 3, No. 3, Report to Congress on Abnormal Occurrences:
July-September, 1980, published in February 1981.
18.
USNRC/AE00, Case Study Report AE00/C204, San Onofre Unit 1 Loss of Salt Water Cooling Event on March 10, 1980, July 1982.
18
f 19.
USNRC Preliminary Notifications PNO-II-82-49 (April 26, 1982) and -43A (May 12, 1982).
20.
Carolina Power and Light Company, Docket No. 50-261, Licensee Event Report 82-03, May 7, 1982.
- 21. USNRC/ Region II, Inspection Report 50-261/82-12, approved May 13, 1982.
22.
USNRC, Preliminary Notification PNO-III-82-052, June 2,1982.
23.
Iowa Electric Light and Power Company, Docket No. 50-331, Licensee Event Report 82-34, June 16, 1982.
24.
USNRC/ Region III, Inspection Report 50-331/82-07 (DPRP), approved August 9, 1982.
These referenced documents are available in the NRC Public Document Room at 1717 H Street, Washington, D.C. 20555, for inspection and/or copying for a fee.
Copies may also be obtained from the editor.
l I
I i
19 m
ABSTRACTS OF OTHER NRC OPERATING EXPERIENCE DOCUMENTS Abnormal Occurrence Reports (NUREG-0090) Issued in May-June 1982 An abnormal occurrence is defined in Section 208 of the Energy Reorganization Act of 1974 as an unscheduled incident or event which the NRC determines is significant from the standpoint of public health or safety.
Under the provi-sions of Section 208, the Office for Analysis and Evaluation of Operational Data reports abnormal occurrences to the public by publishing notices in the Federal Register, and issues quarterly reports of these occurrences to Congress in the NUREG-0900 series of documents.
Also included in the quarterly reports are updates of previously reported abnormal occurrences, and summaries of cer-tain events that may be perceived by the public as significant but do not meet the Section 208 abnormal occurrence criteria.
Date Issued Report 5/82 Report to Congress on Abnormal Occurrences:
October-December, 1981, NUREG-0090, Vol. 4, No. 4 -
During the report period, there were two abnormal occurrences at the nuclear power plants licensed to operate.
One involved a generic concern pertaining to blockage of coolant flow to safety-related syste:as due to aquatic fouling, silt and corrosion pro-ducts.
The other involved seismic design errors at Diablo Canyon Nuclear Power Plant with subsequent suspension of the fuel load and low power operating license for Unit 1.
The report continues to provide update information concerning the accident at Three Mile Island.
In addition, (1) pressurized thermal shock of nuclear reactor nressure vessels, and (2) nuclear power plant construction dd iciencies are discussed as items of interest that do not meet abnormal occurrence criteria.
6/15/82
" Major Deficiencies in Management Controls at a Nuclear Power Plant," Federal Register Notice (47 FR 25793)
This report covers three occurrences of safety concern at Pilgrim Unit 1 which indicated continuing serious deficiencies in management control by the licensee (Boston Edison Company) of certain licensed activities; the Commission determined that this serious deficiency in management should be classified as an abnormal occurrence.
The first item involved failure to comply with provisions of 10 CFR 50.44 regarding the ability to control combustible gas mixtures following postulated accidents.
The second item involved improper maintenance activities which significantly reduced the assurance that certain containment isolation valves 20
Date Issued Report would close when required.
The third item involved operation of the facility with the primary containment dry well temperatures greater than stipulated in the Final Safety Analysis Report which could result in detrimental effects to equipment required to safely shutdown the reactor and to mitigate certain postulated accidents.
The NRC concluded that continued operation of the plant over the long term required significant changes in the control of licensed activities.
Consequently, an Order Modifying License Effective Immediately was issued in regard to such changes, and a proposed civil penalty of $550,000 was issued.
In response, the licensee paid the civil penalty and submitted a comprehensive performance improvement program.
The licensee also restructured corporate functions within its nuclear organization.
In addition, the licensee corrected the identified technical deficiencies.
The NRC is monitoring the licensee's implementa-tion of the improvement program.
21
Bulletins, Circulars,* and Information Notices Issued in May-June 1982 The Office of Inspection and Enforcement periodically issues Bulletins, Circulars and Information Notices to licensees and holders of construction permits.
During the period, two Bulletins and nine Information Notices were issued.
Bulletins are used primarily to communicate with industry on matters of generic importance or serious safety significance; i.e., if an event at one reactor raises the possibility of a serious generic problem, an NRC Bulletin may be issued requesting licensees to take specific actions, and requiring them to submit a written report describing actions taken and other information NRC may need to assess the need for further actions.
A prompt response by affected licensees is req ired and failure to respond appropriately may result in an enforcement action, such as an order for suspension or revocation of a license.
When appropriate, prior to issuing a Bulletin, the NRC may seek comments on the matter from the industry (Atomic Industrial Forum, nuclear steam system suppliers, vendors, etc.), a technique which has proven effective in bringing faster and better responses from licensees.
Bulletins generally require one-time action and reporting.
They are not intended as substitutes for revised license condi-tions or new requirements.
Circulars notify licensees of actions NRC recommends be taken.
Although written responses are not required, the licensees are asked to review the information and implement the recommendations if they are applicable to their facility.
Information Notices are rapid transmittals of information which may not have been completedly analyzed by NRC, but which licensees should know.
They require no acknowledgment or response, but recipients are advised to consider the appli-cability of the information to their facility.
Date Bullet.in Issued Subject 82-01 RI 5/7/82 ALTERATION OF RADIOGRAPHS OF WELDS IN PIPING SUBASSEMBLIES This is a revision to Bulletin 82-01, same subject, March 31, 1982, which discusses the artificial enhancement of the ASME Code specified penetrameter 4T-Hole image in certain piping subassembly shop welds.
The subassemblies were supplied to the Washington Public Power Supply System by the Associated Piping and Engineering Corporation. The NRC requested that certain actions be taken by applicants for operating licenses and holders of construction permits.
Bulletin 82-01 RI updates Table 1, a list of facilities which were supplied the piping subassemblies.
Both bulletins were sent to all holders of a nuclear power reactor operating license or construction permit.
- No circulars have been issued in 1982.
22
Date Bulletin Issue Subject 82-02 6/2/82 DEGRADATION OF THREADED FASTENERS IN THE REACTOR COOLANT PRESSURE BOUNDARY OF PWR PLANTS In the last several years, licensees have reported a signi-ficant incidence of bolts and studs that have failed or become severely degraded because of boric acid corrosion or stress corrosion cracking.
Preliminary results of an NRC staff review of threaded fastener experieiice in nuclear power plants have identified that specific generic actions need to be taken before the study is complete.
This bul-letin requests that certain actions be taken by all pres-surized water nuclear power reactor facilities holding an operating license, and provides information to all other nuclear power reactor facilities holding an operating license or construction permit.
Information Date Notice Issued Subject 82-13 5/10/82 FAILURES OF GENERAL ELECTRIC TYPE HFA RELAYS In 1976, General Electric alerted its customers to the problem of LEXAN and NYLON coil spool surface cracking, and recommended replacement of the cracked spools with Century Series TEFZEL coil spools, or of the entire relay with an HFA Century Series relay.
Recently, there have been several instances of melting of LEXAN and NYLON coil spool material.
In one instance, melted LEXAN material prevented the relay contacts from opening.
This notice was sent to all holders of a nuclear power reactor opera-ting license or construction permit.
82-14 5/12/82 TMI-1 STEAM GENERATOR / REACTOR COOLANT SYSTEM CHEMISTRY /
CORROSION PROBLEM During a long term cold shutdown, radioactivity was detected in the secondary side of both steam generators at Three Mile Island Unit 1.
Examination has revealed corrosion damage to thousands of steam generator tubes and primary to secondary leakage at more than 100 tube locations.
Chemical analysis of the reactor coolant revealed sulfur compounds, which can cause rapid corrosion of some reactor coolant system material, including tubing.
Potential sources of sulfur are being reviewed.
This notice was sent to all holders of a nuclear power reactor operating license or construction permit.
23
i Information Date Notice Issued Subject 82-15 5/28/82 NOTIFICATION OF THE NUCLEAR REGULATORY COMMISSION During nonbusiness hours, each NRC Regional Office diverts its respective telephone numbers to the 24-hour Operations Center in Bethesda, MD, over the Emergency Notification System.
Upon failure of this system, or inability of licensees who do not have this system to reach a Regional Office, the NRC should immediately be called commercially at 202-951-0550.
This notice was sent to all holders of a nuclear power reactor operating license or construction permit.*
82-16 6/28/82 HPCI/RCIC HIGH STEAM FLOW SETPOINTS On February 11, 1982,_the licensee for the FitzPatrick plant (a General Electric boiling water reactor) reported incorrect steam flow differential pressure settings on the high pressure coolant injection (HPCI) system.
The high steam flow trip is provided to detect a break in the HPCI steam supply line and initiate closure of the steam supply isolation valves.
A similar feature exists in the reactor core isolation cooling (RCIC) system.
Although the design calculated setpoints are normally stated in a plant's Final Safety Analysis Report (FSAR), the setpoint values directly depend on the exact configuration of the elbow' tap holes and associated instrumentation.
For this reason, setpoint values are not accurately predictable in all cases and must be established by actual testing.
General Electric recommended that its customers verify that their FSAR-defined HPCI and RCIC steam flow setpoints are consistent with startup test data.
This notice was sent to all holders of a nuclear power reactor operating license or construction permit.
82-17 6/11/82 OVERPRESSURIZATION OF REACTOR COOLANT SYSTEM On November 28 and 29, 1981, the Turkey Point Unit 4 reac-tor coolant system (RC5) was overpressurized during start-up following a refueling outage.
At the North Anna plant, from May 19-22, 1982, the overpressure protection system was inoperable.
Each of these events involved failure of two redundant systems designed to provide overpressure protection.
The concern is that without prompt operator action, such failures increase the potential for brittle fracture of the reactor pressure vessel from overstress during pressure transients.
This notice was sent to all
" Calls normally directed to the Region I office only (King of Prussia, Pennsylvania) after working hours should be made to the NRC Headquarters operator on 800-368-5642.
24
Information Date Notice Issued Subject holders of a nuclear power reactor operating license or construction permit.
82-18 6/11/82 ASSESSMENT OF INTAKES OF RADI0 ACTIVE MATERIAL BY WORKERS Licensees must demonstrate that their methods for assessing intakes of radioactive material by workers complies with ICRP-2 (Report of the ICRP Committee II on Permissible Dose for Internal Radiation, 1979 revision) methodology, until the revision of 10 CFR 20 incorporating new ICRP methods and recommendations is adopted.
This notice was sent to all holders of a nuclear power reactor operating license or construction permit, research and test reactors, fuel facilities, and Priority I material licensees.
82-19 6/18/82 LOSS OF HIGH HEAD SAFETY INJECTION EMERGENCY B0 RATION AND REACTOR COOLANT MAKEUP CAPABILITY On February 12, 1982, McGuire Unit 1 experienced a loss of high head safety injection emergency boration and reac-tor coolant makeup capability.
Hydrogen from the positive displacement pump (PDP) suction dampener entered the com-mon suction of the chargi7g system causing both centrifugal charging pumps (CCPs) and the PDP to be inoperable.
Fail-ure to properly vent the PDP and its associated piping prior to opening the valve that isolates the equipment from the CCPs may have contributed to the event.
There is a concern that a potential single component malfunction, personnel error, or maintenance error in the safety or nonsafety-related portion of the system could lead to the same consequences, even for a system of a different design.
This notice was sent to all holders of a nuclear power reactor operating license or a construction permit.
82-20 6/28/82 CHECK VALVE PROBLEMS A number of problems were recently reported involving swing check valves supplied by two manufacturers:
Alloy Steel Products Company (ALOYCO) and Pacific.
At the Palisades plant, where such cneck valves separate the low and high pressure safety injection systems, abnormal wear of the valves was discovered during modifications.
Fail-ure of two in-series check valves to function as a pres-sure isolation barrier could cause an overpressurization and rupture of the low pressure system.
This would result in a loss-of-coolant accident (LOCA) that bypasses contain-ment and renders inoperable some of the equipment needed to mitigate a LOCA.
At the Susquehanna Steam Electric Station Unit 1, three problems with Pacific check valves 25
Information Date Notice Issued Subject were reported:
(1) disk assembly to body interference and excessive packing friction, (2) excessive wear at the hinge arm / disk stud interface, and (3) disk stud breakage.
These check valves are used in many nonsafety systems, as well as safety systems such as the residual heat removal, reac-tor core isolation cooling, and core spray systems.
This notice was sent to all holders of a nuclear power reactor operating license or construction permit.
82-21 6/30/82 BUILDUP 0F ENRICHED URANIUM IN EFFLUENT TREATMENT TANKS This notice informs about the potential buildup of enriched uranium in large liquid waste tanks after waste solutions have been released from criticality control.
For nuclear criticality control purposes, licensees must sample or monitor effluent streams containing uranium prior to re-lease of such liquids from favorable geometry hold systems.
On two occasions significant quantities of uranium were detected in large, nonfavorable geometry effluent treatment systems as a result of post precipitation or other chemical /
mechanical action.
This information should be brought to the attention of all personnel having operational and safety responsibilities for waste treatment systems.
The notice was sent to all uranium and plutonium fuel fabrication licensees.
26
i Operating Reactor Event Memoranda Issued In May-June 1982 The Director, Division of Licensing, Office of Nuclear Reactor Regulation (NRR),
disseminates information to the directors of the other divisions and program offices within NRR via the operating reactor event memor;.ndum (OREM) system.
The OREM documents a statement of the problem, background information, the safety significance, and short and long term actions (taken and planned).
Copies of OREMs are also sent to the Offices for Analysis and Evaluation of Operational Data, and of Inspection and Enforcement for their information.
Date OREM Issued Subject 82-01 5/25/82 WATER HAMMER EVENTS AT MILLSTONE UNIT 1 Introduction of water into the isolation condenser steam supply line during the August 10, 1981 overfilling event et Millstone Unit I caused the isolation condenser to be unavail-able when required.
Later, when the isolation condenser was being 71 aced in the stand-by condition, a water hammer occurred causing minor damage to the concrete expansion bolts associated with the isolation condenser steam supply line restraints.
The concerns raised by water hammer events that are caused by vessel overfill are (1) potential for damage to a safety sys-tem, and (2) potential for damage to steamlines or safety relief valves.
Short term actions taken by the licensee included adjustments to the pressura control system, the feedwater controller and regulating valve, and acoustic monitors.
Long term actions include implementing a feedwater pump trip on high reactor water level (high level setpoint will be below tne bottom of the inlet nozzle to the isolation condenser, but above the normal water level), lowering the reactor water cleanup system isolation setpoint, and modifying the~10w flow feedwater con-trol system to limit feedwater nozzle cracking.
The licensee is also evaluating a design change to determine the ability to prevent raactor vessel overfill following a scram.
The NRC staff is currently reviewing water hammer events generally as Unresolved Safety Issce USI A-1.
Data from this study shows there have been no other water hammer events in BWR isolation condensers.
In addition, all BWRs except Nine Mile Point and Oyster Creek (whose plant-specific designs have been reviewed and found acceptable) have a vessel high water level trip to preclude entrainment of liquid in the isolation condenser steam supply.
27
Engineerina Evaluations and Case Studies Issued in May-June 1982 The Office for Analysis and Evaluation of Operational Data (AE00) has as a primary responsibility the task of revewing the operational experience reported by NRC nuclear power plant licensees.
As part of fulfilling this task, it selects events of apparent interest to safety for further review as either an engineering evaluation or a case study.
An engineering evaluation is usually an immediate, general consideration to assess whether or not a more detailed, protracted case study is needed.
The results are generally short reports, and i
the effort involved usually is a few staffdays of investigative time.
i Case studies are in-depth investigations of appareatly significant events or situations. They involve several staffmonths of engineering effort, and result in a formal report identifying the specific safety problems (actual or potential) illustrated by the event and recommending actions to improve safety and prevent recurrence of the event.
Before issuance, this report is sent for review and comment to the applicable utility, the Institute of Nuclear Power Operations, the Nuclear Safety Analysis Center, and other NRC offices.
These AE00 reports are made available for information purposes and do not impose any requirements on licensees.
The findings and recommendations contained in these reports are provided in support of other ongoing NRC activities concerning.the operational event (s) discussed, and do not represent the position or requirements of the responsible NRC program office.
Case Date Study Issued Sub.iect C203 5/82 Survey of Valve Operator-Related Events Occurrina Durina 1978, 1979, and 1980 This survey report on valve operator-related events provides:
(1) a brief summary of related previous reviews and recommenda-tions developed by both the NRC and industry groups; (2) a review and evaluation of events that occurred during 1978, 1979, and 1980; and (3) recommendations that would lead to appropriate definition and/or resolution of problems discerni-ble from operating event experience during the three year time span of the survey.
The primary source of information was licensee event reports (LERs).
The survey of LERs provided sufficient information to indicate that motor operator-related events are the greatest single categcry of reported valve operator events.
Further investi-gation revealed that events could be grouped into three major problem categories involving torque switches, limit switches, and motors.
28
r --
s f
f
,./
')
Engineering - Dhe Evaluation Iss'ued Subject y
E223 5/11/82 Ina{Nettent to5s of coolant Events at Sequoyah Nuclear Power Planti Units 1 and 2 4
Separate events at Unit 1 and Unit 2 at,tne Sequo'yab.
nuclear plant resulted in the inadvertent partial loss of reactor coolant during shutdown cooling.
The safety significance concerns the loss-of-coolant event inside containment which could result in the loss of the one i
c train of the decay heat removal and emergency cor'e ' tool-1 ing recirculation system required for mitigation of such an event.
Changes were jecommended to improve quality i
assuranch, valve,t.ontrcF, and the procedures and train-i e
ing regarding prevention and mitigation of such events.
s j
s s E224 1
5/21/82 Generic Concerns Asscciated with the Ginna Steam.
Generator Tube Rupture T ier.t
( [!
This evaluation presents a list of generic concerri #
'y s
associated with the January 25,1$82 steam generator i tube rupture at Ginna.
These findings and recommenda-tions sere based on AE0D's participation.in the prebara-tienofNUREG-0909,NRCReportontheJanuary25,1Q2 Steam Generator Tube Rupture at R.
E". Ginna Nuclear */
Power Plant.
I E225 6/1/82 Degradatin'ofBWRScramPilotSolenhdValvesDueto 7
Abnorma) Power S'ipply Voltage In August 1981 during control rod drive testing at Grand Gulf Nuclear' Station Unit 1, several scram pilot scle-noid valves were found stuck in the energized position when the solenoids were deenergized.
The affected scram inlet and outlet valves did not actuate.
Apparently, the valves were damaged by being operated with insuffi-cient voltage to the salenoid coils from the power supply ferjthe reactor protection system due to insuffi-cient cable size.
Voltage measurements for line losses, optimally adjustec protective circuits, functional tests and better surveillance and equipment replacement
+
schedule < ard recommended.
)
l E226 6/18/82 Inopera'bilitlyofInstrumentationDuetoExtremeCold Weather
/
Anumberofeventsinvolvingfrozenlinesoninstrument i
sensing and sampling lines led to recommendations to l
improve NRC's inspection procedure and to better inform
/
licensees of the broad experience.
- /
,/
l 29
i i
l Engineering Date Evaluation Issued Subject E227 6/24/82 Failure of Engineered Safety Features Manual Initiation Pushbutton Switches This report covered a December 12, 1981 event at the McGuire Unit 1 nuclear station, where the normal switches failed to actuate all their devices during a periodic test. 'When actuated, some contacts changed state while others did not due to a silver sulfide coating building I
up on the silver plated contacts.
AE00 believes this concern could apply to similar types of safety-related switches in low voltage and low current applications and recommended that licensees be made aware of the problem.
E228 6/25/82 Hopetitive Overspeed Trips of the Steam-Driven Emergency Feldwater Pump on Initial Start at Arkansas Nuclear One, Unit 2 This report covered a review of the history of the repetitive overspeed trips of the steam-driven emergency feedwater pump on initial startup at Arkansas Nuclear One, Unit 2.
Condensate in the steam supply line was initially thought to be the cause, but further study indicated that the failures were occurring after a long,
/~
idle period (greater than 40 hours4.62963e-4 days <br />0.0111 hours <br />6.613757e-5 weeks <br />1.522e-5 months <br />).
Following these efforts the control oil tube was considered.
The design specifications called for 3/8" 0.D. tubing with a 0.035" 7
wall thickness.
The originally installed 3/8" tubing was 0.065" wall thickness which resulted in a 35% reduc-tion in available flow area for control oil and caused sluggish response of the governor valve.
The control oil tube was replaced.
There has been no recurrence of the failures to date.
+
,96 s
These documents are available in the NRC Public Document Room at 1717 H Street, Washington, D.C. 20555, for inspection and/or copying for a fee.
Copies may also be obtained from the editor.
l 30
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