ML20087K176
| ML20087K176 | |
| Person / Time | |
|---|---|
| Issue date: | 02/29/1984 |
| From: | Massaro S NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-BR-0051, NUREG-BR-0051-V05-N5, NUREG-BR-51, NUREG-BR-51-V5-N5, NUDOCS 8403260056 | |
| Download: ML20087K176 (37) | |
Text
NUREGlBR-0051 Vol. 5, No. 5
!O POWER REACTOR EVENTS k..... /
United States Nuclear Regulatory Commission Date Published: February 1984 Power Reactor Events is a bi-monthly newsletter that compil'es 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 from operational experience to the various plant personnel, i.e.. rnanagers, 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) 492-9530 or at Mail Stop 016, Washington. DC 20555.
Table of Contents Page 1.0 SUMMARIES OF EVENTS 1.1 Main Steam Isolation Valve Failures......................................................
1 1.2 Reactor Depressurization Resulting from Faulty Safety Relief Valve.................
2 1.3 Inadvertent Drainage from Reactor Vessel.........
4 1.4 L oad Rejection ftom 75% Po wer............................................................
5 1.5 Inadvertent Isolation of Emergency Bus............
7 1.6 Fuel Line Rupture Resulting in Fire........................................
8 1.7 Improperly Installed Fire Dampers........
9 1.8 Inoperability of High Pressure Coolant Injection.......................................
10 1.9 References................
12 2.0 ABSTRA CTS OF O THER NRC OPERA TING EXPERIENCE DOCUMENTS 2.1 Abnormal Occurrence Reports (NUREG-0090)...............................................
13 2.2 Bulletins and information Notices....
14 2.3 Case Studies and Engineering Evaluations....................................................
22 2.4 Generic L e t ters...........................................................................
28 2.5 Operating Reactor Event Memoranda..........................................................
30 2.6 Regulatory and Technical Reports (NUREG-0304)....................................
31 Editor: Sheryl A. Massaro B403260056 B40229 Associate Editor: Steven E. Trenery PDR NUREG Office for Analysis and Evaluation BR-0051 R PDR of Operational Data U.S. Nuclear Regulatory Commission Pariod Covered: September-October 1983 Washington, D.C. 20555
1.0 SUMMARIES OF EVENTS 1.1 Main Steam Isolation Valve Failures On September 17, 1983, with Farley Unit 2* in hot standby, the licensee was sicuring steam to the turbine building as part of a normal shutdown.
At 4:35 p.m., the loop 2 upstream, loop 2 downstream, and loop 3 upstream main steam isolation valves (MSIVs) initially failed to close under no flow condi-tions.
Both upstream MSIVs were subsequently closed, but the loop 2 down-stream MSIV could not be closed with manual assistance and was declared inoperable.
Two apparently unrelated causes for these failures have been identified.
The upstream MSIVs may have initially failed to close due to binding in the valve packing.
The downstream MSIV failed to close due to separation of the valve disk from the disk arm.
Subsequent to the upstream MSIV failures, the valve test cylinders were actuated to provide air assisted motion.
This motion was sufficient to free the upstream valves, which were closed at about 5:15 p.m.
Inspection of the valve packing revealed no apparent hardening or other degrada-tion.
However, further investigation revealed a difference between the packing used on Unit 1 and Unit 2.
The significance of this difference has not yet been determined.
To preclude recurrence of this problem, the licensee will (1) inspect and replace the packing on all MSIVs during each refueling outage in accordance with the latest vendor recommendations, and (2) stroke the MSIVs to ensure free movement after packing adjustments during heatup.
These actions will be continued until the licensee and the MSIV vendor determine that the problem has been eliminated or a design change has been made to preclude the need for continued corrective actions.
In addition, failure of all four disk fasteners on the downstream MSIV resulted in the valve disk becoming separated from the disk arm. These valves are de-signed with Belleville washers to absorb the impact of MSIV closure under steam flow conditions and ensure proper disk-to-seat alignment.
This design appar-ently allowed flexure during flow conditions while the valve was on the open stop.
It is believed that this flexure exposed the fasteners to cyclic loads which resulted in excessive thread wear and eventual failure.
Excessive system vibration, noted during operation, may have exacerbated this flexure problem.
In addition to the failed fasteners on the downstream MSIV, at least one fastener was found loose on each of the five remaining MSIVs (a total of eight additional fasteners were loose).
A design change being implemented to pre-l clude recurrence of this problem consists of:
Adding two open stops to the body of each MSIV.
These new stops and the existing stop will restrain the valve disk while in the normally open position, thus eliminating the flexure motion of the disk.
- Farley Unit 2 is an 814 MWe (net) PWR located 28 miles southeast of Dothan, Alabama, and is operated by Alabama Power.
1
Replacing the four studs used to fasten the disk to the disk arm with studs that have interference fit threads.
This design will eliminate the need for the fastener studs to rely on the Belleville washers to maintain preload.
This change is to eliminate the excessive wear of the threads due to disk movement.
1 Utilizing an interference fit between the fastener stud hex nut and the stud and installing a set screw through the pitch diameter of the stud and nut.
This will assure constant preload on the disk assembly via the Belleville washers.
In addition, the valve manufacturer has notified customers about the problem and has recommended that the valves be inspected.
The excessive vibration noted during operation at Farley well be monitored again after startup.
- Also, plant operators, shift supervisors, and shift technical advisors have been instructed to actuate the valve test cylinders should an MSIV fail to close when required.
(Ref. 1.)
1.2 Reactor Depressurization Resulting from Faulty Safety Relief Valve On September 15, 1983, Quad Cities Unit 1* was being shut down for planned maintenance and surveillance activity.
After the generator was taken off line, and with the reactor at about 20% rated thermal power, the licensee began the required routine surveillance test of the three-stage Target Rock safety relief valves.
The test began at a reactor pressure of 960 psig.
At 12:26 a.m., indications showed that the valve had opened and had begun to reseat, but did not reseat fully.
Reactor pressure was reduced to 360 psig before the valve closed at 12:40 a.m.
The reactor was manually scrammed during attempts to reseat the valve.
During the event, reactor cooling water temperature decreased about 85 F in 20 minutes (the technical specification limit is 100 F per hour), and the pres-sure suppression pool, which quenches the blowdown steam, reached 101 F (tech-nical specifications require shutdown at 110*F).
The valve was removed and inspected.
It was found that the second stage pilot seat (see Figure 1) had eroded slightly, allowing the steam above the piston to escape, permitting the valve to remain open. The safety relief valve (Target Rock,.model 67F) was replaced and tested successfully.
(Refs. 2 and 3.)
- Quad Cities Unit 1 is an 810 MWe (net) BWR located 20 miles northeast of Moline, Illinois, and is operated by Commonwealth Edison.
2
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- 1. 3 Inadvertent Drainage from Reactor Vessel 2
On Septemher 14, 1983, with LaSalle Unit 1* in cold shutdown, between 5,000 i
and 10,000 gallons of water were inadvertently drained from the reactor vessel.
The fuel remained covered at all times.
The event occurred when a reactor operator opened the low pressure coolant injection (LPCI) valve in accordance with the approved surveillance procedure.
As called for by the procedure, the LPCI recirculation valve to the suppression pool and the containment spray valve were open.
The static level of the water in the reactor vessel caused water to drain from the reactor vessel past a failed check valve to the sup-pression pool and out the lower level containment spray nozzles.
The drainage occurred for approximately three minutes before it was terminated by a combina-tion of operator action and automatic containment isolation due to low vessel water level.
A residual heat removal (RHR) system relay logic test had been in progress.
Just before the event, the B RHR loop was lined up to support this test.
This lineup called for both drywell spray valves, the suppression pool spray valves, the test return valve to the suppression pool, and the C RHR loop injection valves to be opened.
The procedure then required opening the B RHR loop injec-tion valve, which left the injection check valve as the only isolation valve for reactor water level. When the B RHR loop injection valve was opened, a rapid decrease in reactor water level was noted (from +50 inches to 0 inch as indicated on the upset range recorder).
The level decrease was stopped by j
closing the B RHR loop injection valve.
Most of the water lost from the i
reactor vessel drained to the suppression pocl.
The remainder drained to the drywell.
The cause of the event was determined to be the B LPCI check valve sticking in j
the open position.
Inspection of the valve operator revealed that the valve had been timed improperly during earlier maintenance on it.
Each interfacing
]
gear between the check valve and the air operator has a timing mark used to I
align the gears for proper reassembly after maintenance.
The timing mark on the spline shaft of the check valve had been confused with a scar mark which j
was used for alignment.
This aligned the check valve and the air operator such that the check valve was left open about 35* with the air operator inhibiting movement in the closed direction.
A second cause for the event became apparent when, after proper timing, the check valve did not pass the local leak rate test.
Further inspection showed that the packing gland was too tight on the check valve shaft and would not allow free rotation to permit full closure.
l Actions taken by the licensee as a result of this event included the following:
i l
The B LPCI check valve was properly timed, its packing gland was adjusted to permit free movement, and the subsequent local leak rate test was l
successfully completed.
l
- LaSalle Unit 1 is a 1078 MWe (net) BWR located 11 miles southeast of Ottawa, Illinois, and is operated by Commonwealth Edison.
4
The timing mark on the spline shaft of the 8 LPCI check valve was more clearly identified to prevent confusion.
Several other testable check valves were inspected for proper timing and for clearly indicated timing marks.
None of these check valves were timed improperly, nor were their timing marks unclear.
Maintenance performed on these testable check valves was reviewed back to the last successful local leak rate test of each valve, to determine which check valves had packing gland adjustments in the interim.
It was found that adjustments had been performed on two valves without a subse-quent local leak rate test.
This testing was then successfully performed on the valves.
Procedural changes include (1) a requirement to perform a local leak test whenever maintenance is performed on the packing gland of a testable check valve; and (2) a requirement that during RHR system relay logic testing the manual RHR injection stop valve be closed on the RHR loop being tested.
(Refs. 4 and 5.)
1.4 Load Rejection from 75% Power On September 12, 1983, Summer Unit 1* underwent a complete loss of main turbine generator load.
Prior to the event, the plant was at 75% power.
The reactor rod control system was in manual while other control systems (steam generator level, pressurizer level, and pressure) were in automatic control.
In order to conduct preventive maintenance on balance-of plant inverter 5906, it was necessary to shift APN-5906, a balance-of plant 120 V ac panel, from its normal inverter power source to its backup power supply, APN1FX.
This evolution was discussed with the control room personnel who were aware that during the transfer, which is a " dead bus" transfer, several alarms and indicators would spike.
The shift supervisor, control room foreman, and the reactor operator were present in the main control room.
All of these individuals are licensed as senior reactor operators.
No other evolutions were in progress at the time, and the power escalation was on hold.
APN-5906 provides normal control power to the turbine electrohydraulic control (EHC) unit.
The turbine permanent magnet generator (PMG) provides backup power to the EHC unit such that on loss of the normal power supply, it will assume the load through electronic auctioneered logic.
At 8:18 a.m., APN-5906 was transferred from its normal to its backup power source.
This operation involves opening the normal incoming breaker, repositioning a mechanical interlock, and closing the backup incoming breaker.
It takes approximately 1 to 5 seconds to perform this operation.
When APN-5906 was deenergized, the normal -22 V dc power source to the EHC unit was lost and the turbine PMG -22 V de power source malfunctioned.
Due to the complete momentary loss of the -22 V dc control power, the turbine control and intercept valves closed, resulting in a total turbine generator load reduction.
CSummer Unit 1 is a 900 MWe (net) PWR located 26 miles northwest of Columbia, South Carolina, and is operated by South Carolina Electric and Gas.
5
With the reactor rod control system in manual, the reactor power remained relatively constant.
Reactor coolant system (RCS) temperature started to increase, with pressurizer level and pressure increasing accordingly.
In the control room, the operators received several annunciator panel alarms.
Included was the. entire reactor trip first out alarm panel, with the "first out" indicating pressurizer high pressure as the first event.
Since the reactor operator was expecting to receive alarms as a result of the APN-5906 transfer, he verified, by the protection bistable monitor lights and the pro-cess instrumentation, that a reactor trip condition did not exist.
He observed that RCS temperature, pressure and pressurizer level were increasing, and in-formed the shift supervisor.
Between 8:18 and 8:19 a.m., the RCS temperature (T-Ave) exceeded the maximum value as specified in technical specifications (592 F) by approximately 2.5 F.
This caused the licensee to enter the Action Statement of the technical specifications.
There were no adverse consequences because T-Ave was restored below 592"F within approximately 3 minutes.
The increase in T-Ave and the decrease in reference temperature (T-Ref), due to a decrease in the turbine first stage pressure, caused the steam dump system to arm.
Four of the eight condenser steam dump valves opened as required.
(Two of these valves had been isolated prior to this event for maintenance.)
Atmospheric steam dump valves also failed to operate as required.
The three main steam power operated relief valves (PORVs) did operate.
This resulted in approximately 50% of the steam dump system opening, which was insufficient to reduce T-Ave.
Main steam pressure increased to approximately 1195 psig, result-ing in the opening of several main steam line safety valves.
This total steam demand terminated the increase in RCS T-Ave.
Pressurizer pressure increased to approximately 2359 psig, resulting in the actuation of pressurizer PORV PCV-445B (setpoint of 2335 psig) to relieve pressurizer pressure.
The remaining two pressurizer PORVs had been manually isolated prior to this event.
The shift supervisor observed that the steam dumps actuated, that the generator load was 0 MW, that the generator breaker was not open, and that the turbine was not tripped.
He evaluated the load rejection as being associated with the APN-5906 transfer and initiated action to stabilize the plant.
The reactor operator commenced emergency boration and manual control rod insertion to reduce T-Ave.
The control room foreman started running the load reference back to zero as required by the operating procedure for load rejections.
When APN-5906 was reenergized, the -22 V dc EHC control power was restored.
At 8:21 a.m., the turbine control and intercept valves automatically reset and opened to reload the turbine generator to the value set on the load reference (243 MW).
The steam demand was now sufficient to reduce T-Ave to 583 F and to reset the main steam safety valves.
A T-Ave versus T-Ref mismatch still existed, and the steam dump to condenser valves were still open.
The shift supervisor coordinated with the system load dispatcher to continue a rapid load. increase in order to close the remaining steam dump valves and match T-Ref to T-Ave.
The reactor operator terminated emergency boration after having inserted approximately 150 to 200 gallons of 7% boric acid.
Between 8:21 and 8:30 a.m.,
the main generator was loaded to approximately 570 MW, thus closing all steam dump valves.
At this point, recovery from the transient was essentially complete with the operators continuing to stabilize the plant utilizing normal operating procedures.
6
Following the event, the atmospheric steam dump valves and condenser steam dump-valves were isolated and functionally tested, and the -22 V de PMG power supply was repaired.
Inverter 5906 was returned to service on September 17.
Additional corrective actions planned as a. result of the event are as follows:
Determine if PMG power supplies and auctioneered outputs can be tested on line to increase reliability.
Evaluate procedures for improvement in removal and restoration of APN-5906 and other auxiliary electrical supply panels.
Restrict preventive maintenance tasks to essential equipment and allow these activities to be performed only with procedures that have received detailed evaluation or have been proven during off line conditions whenever possible.
Evaluate the need for additional operator guidance for recovery from runback or load rejection events.
Perform a reevaluation of NRC's IE Bulletin 79-27, " Loss of Non-class 1E Instrumentation and Control Power System Bus During Operation," issued October 25, 1979, to determine if differences exist from the original evaluation.
(Refs. 6 and 7.)
- 1. 5 Inadvertent Isolation of Emergency Bus At Beaver Valley Unit 1* on May 24, 1983, during normal operation at 99% power, relay testing was being performed on the 18 system station service transformer (relay A).
Spurious actuation of relay B occurred, causing inadvertent isola-tion of the IDF emergency bus from its normal supply.
Diesel generator No. 2 automatically started and picked up the bus, sequencing on all appropriate loads.
An investigation into the cause of this event has revealed a number of contrib-uting factors.
Lockout relay A has a normally open contact which closes to energize relay B.
Relay A cannot be cleared by itself without isolating a sub-stantial portion of the protection circuitry for the IB system station service transformer.
The relay was therefore tested uncleared.
This should have posed no problem since relay A is a de relay and should not operate on the low ac test voltages applied to the contacts.
Unknown to the test crew, however, there was a' capacitor installed between the 125 V de control power lines and ground for su.rge suppression.
This suppression capacitor inadvertently dis-charged when the test signal was applied to the relay contacts and activated relay B.
The capacitor discharged because the test equipment utilized a grounded ac source.
This ground provided a potential path for current flow, and allowed discharging of the. capacitor to occur.
When relay B activated, the feeder breaker to the emergency bus tripped open.
This action caused a " dead bus" condition on the emergency 4 kV bus, which
~" Beaver Valley Unit 1Lis an 810 MWe (net) PWR located in Pennsylvania, five
'idiles east of East Liverpool, Ohio, and is operated by Duquesne Light.
I 7
i intiated an automatic start of the diesel generator.
After the diesel generator attained design output vc,1tage, the output breaker closed and sequencing of all appropriate loads occurred.
The test which led to the spurious activation was rerun.
During subsequent performance of the relay test, relay B activated each time, as it had in the initial event.
This verified that it was the test that caused the relay actu-ation, and not the shorting of contacts or carelessness on the part of the technician.
Relay personnel at Beaver Valley will be reinstructed in the i
performance of tests en circuits such as this and others which require special test considerations.
(Ref. 8.)
1.5 Fuel Line Rupture Resulting in Fire At Grand Gulf Unit 1* on September 4, 1983, a fuel line ruptured on the No. 11 diesel generator.
Eight personnel working in the area immediately evacuated.
The engine was manually stopped and the outside fresh air fans were secured when a fire was reported at the engine.
The fire brigade responded with fire hoses and portable extinguishers.
It was noted, however, that the automatic fire u ter deluge valve had not opened.
The manual release was ineffective since the drop weight had already been released by the automatic function.
A mechanic was able to open the valve by removing the actuator enclosure box cover and striking the top of the weight, and the fire was extinguished.
Initial inspection of the diesel engine revealed that the main fuel supply tubing which delivers fuel oil from the engine driven fuel oil booster pump to the left and right bank fuel headers had separated at a tee connection.
The separation resulted in fuel oil spraying on a hot exhaust manifold, entering the left bank turbocharger and igniting. The fuel oil flow continued to feed the fire until the engine reached a complete stop and the headers were drained.
The separation of the main fuel supply tubing was caused by a crack that occur-red at the ferrule of the fitting used to connect the tubing to the fuel header tee.
A metallurgical evaluation of the failed tubing indicated that the failure resulted from very high cycle fatigue.
The high cycle fatigue resulted because a vendor-supplied support immediately downstream of the fitting was not provided.
A support bracket was added to the fuel oil header.
Components which were located in the fire area were replaced since the ability to carry out their design function was in question.
Other components which may have been sub-jected to heat or water damage were inspected and either replaced or reworked, depending on their condition.
Any item whose condition could not be accurately evaluated was replaced.
Components replaced included 41 items of the engine mounted equipment, 24 items of engine mounted instrumentation, 15 items of engine mounted electrical equipment, and 15 miscellaneous items.
After all work had been completed, the diesel generator was subjected to a maintenance run.
During the maintenance run, the engine was instrumented for vibratory analysis.
The preliminary results of this analysis revealed that the
- Grand Gulf Unit.1 is a BWR with a design electrical rating of 1250 MWe (net),
and was granted a-low power license in June 1982.
It is located 25 miles south of Vicksburg, Mississippi, and is operated by Mississippi Power and Light.
8
cngine exhibited vibrations that were well within the acceptable limits for this type of machinery.
The failed fire water deluge valve was a 6-inch Model C valve, serial number S10774, manufactured by Automatic Sprinkler Corporation of America.
When the trip signal was received from the local control panel at the time of the event, the drop weight was released but the valve failed to open.
The valve and the release mechanism were tested and components were removed and examined.
No significant abnormal conditions were noted.
Although some excessive friction was noted between the weight and weight guide rod, the valve operated properly during subsequent testing.
Corrective actions included reworking of the weight's upper collar guide, and lubrication of the rod, latch hinge pin, and clapper hinge pins.
The surveillance procedure also has been revised to visually verify that the clapper has lifted and locked open following the test under normal system pressure.
(Ref. 9.)
- 1. 7 Improperly Installed Fire Dampers On June 13, 1983, at Crystal River Unit 3,* while moving duct work to accom-modate a plant modification, licensee personnel discovered that a required fire damper had not been installed in a ventilation duct between the auxiliary building and the control building.
In evaluating and implementing corrective actions, the licensee identified two additional problems.
The design drawings call for a 3-hour fire rating on the dampers, but the procurement documents specified only a 1-1/2-hour rating.
Thus, 56 dampers installed in safety-related areas did not have the required fire rating.
Further investigation revealed that 21 of the 56 identified fire dampers were not properly located within the duct system.
These dampers were installed within the ducts in the fire area and not within the fire wall penetration as required by the Design and Installation Code, National Fire Protection Association Standard 90A (NFPA-90A), Air Conditioning and Ventilating Systems.
Redundant safety-related equipment and components at nuclear power plants are required to be separated by distance or by fire-resistant walls, floors, cnclosures, or o'.her types of fire barricrs.
All penetrations in the fire barriers are required to be protected against the spread of fire in order for the barriers to be effective.
Ventilation duct penetrations of fire barriers are required to be protected by means of fire dampers which are arranged to close in the event of fire.
If these dampers are installed improperly, it is possible that fire could pass through a fire barrier and thereby jeopardize redundant safety-related systems.
f The situation at Crystal River Unit 3 may also exist at other facilities.
l During recent NRC investigations, improperly installed fire dampers have in fact been identified at Farley, Grand Gulf, Ft. St. Vrain, Vermont Yankee, and CCrystal River Unit 3 is a 782 MWe (net) PWR located seven miles northwest of Crystal River, Florida, and is operated by Florida Power.
l l
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Davis Besse Unit 1.*
Discrepancies at these plants exist primarily because of inadequate design data or because the design documents did not conform to the criteria of NPFA-90A or of the damper manufacturer.
On October 21, 1983, the NRC issued Inspection and Enforcement Information Notice 83-69, 4mproperly Installed Fire Dampers at Nuclear Power Plants," to encourage licensees to review fire damper installations at their facilities and determine whether (1) the correct dampers are installed, and (2) the damper installation is in accordance with relevant criteria.
(Refs. 10 and 11.)
1.8 Inoperability of High Pressure Coolant Injection At Pilgrim ** on September 29, 1983, the low pressure section of the high pressure coolant injection (HPCI) suction piping became overpressurized at 11:15 a.m.,
during functional testing of the HPCI system logic.
The licensee has attributed the cause of this event to personnel error while conducting more than one surveillance test at the same time and not ensuring that prerequisites and initial conditions of all steps were met.
This led to the simultaneous opening of two H'PCI discharge valves (2301-8, -9) and, with a partially opened testable injection check valve (2301-7), a sudden pressuriza-tion spike in the reverse direction through the HPCI pump.
After the initial identification of this event, control room operators opened the HPCI test line to the condensate storage tank and shut valve 2301-8 to depressurize the system.
The control room operators received a high suction pressure alarm, a lube oil high temperature alarm, various smoke detector alarms and a 250 V dc battery bus ground.
Local investigation revealed water and oil on the floor of the HPCI quadrant.
- Farley Unit 1 is an 804 MWe (net) PWR and Unit 2 is an 814 MWe (net) PWR located 28 miles southeast of Dothan, Alabama.
They are operated by Alabama Power and Light.
Grand Gulf Unit 1 is a BWR with a design electrical rating of 1250 MWe (net),
and was granted a low power license in June 1982.
It is located 25 miles south of Vicksburg, Mississippi, and is operated by Mississippi Power and Light.
McGuire Unit 1 is a 1180 MWe (net) PWR located 17 miles north of Charlotte, North Carolina, and is operated by Duke Power.
Ft. St. Vrain is a 330 MWe (net) HTGR located 35 miles north of Denver, Colorado, and is operated by Public Service of Colorado.
Vermont Yankee is a 504 MWe (net) BWR located five miles south of Brattleboro, Vermont, and is operated by Vermont Yankee.
Davis Besse Unit 1 is an 874 MWe (net) PWR located 21 miles east of Toledo, Ohio, and is operated by Toledo Edison.
- Pilgrim is a 670 MWe (net) BWR located four miles southeast of Plymouth, Massachusetts, and is operated by Boston Edison.
10
The control room operators attempted to perform an operability test to deter-mine the exact status of the HPCI system.
After starting the auxiliary oil pump and opening the turbine stop and control valves, the stop valve would not trip from the control room.
At 12:50 p.m. on September 29, the HPCI system was declared inoperable and alternate testing required by the technical specifica-tions was begun.
Findings and corrective actions by the licensee included the following:
An inspection of major components, piping, and supports was performed.
The smoke alarms were found to have been initiated by vapors from heated sections of non-lagged HPCI suction piping and the battery ground caused by water from a ruptured gland seal condenser gasket spraying a limit switch.
The testable injection check valve (2301-7) was manually operated and was found to be free to move fully closed and open.
However, the valve stem was rusted to a part of the actuating device.
This condition was deter-mined to have possibly held the check valve disk in a partially open posi-tion during normal operations with no differential pressure across it, but not to have caused the valve to bind up.
This corrosion problem was repaired.
In addition, the valve was verified to be intact and leak tight following a test with downstream vent and test connections.
A preliminary analysis by a consultant concluded that the stresces did not exceed yield stress and that the low pressure piping was operable.
The HPCI system was demonstrated operable during a full flow recirculation test and an injection test into the reactor vessel.
The licensee has initiated a transient analysis to determine whether any actions are needed to ensure long term acceptability of the low pressure piping.
The licensee has also held a critique with operators and instrumentation and control technicians, and has initiated administrative actions to ensure strict compliance with surveillance procedure action steps.
(Refs 12 through 14.)
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i
- 1. 9 References (1.1) 1.
Alabama Power Company, Docket 50-364, Licensee Event Report 83-39, October 19, 1983.
(1.2) 2.
NRC, Preliminary Notification PNO-III-83-86, September 15, 1983.
3.
Commonwealth Edison Company, Docket 50-254, Licensee Event Report 83-35, October 4, 1983.
(1.3) 4.
NRC, Items of Interest for Week Ending September 23, 1983,.
5.
Commonwealth Edison Company, Docket 50-373, Licensee Event Report 83-105, September 27, 1983.
(1.4) 6.
South Carolina Electric and Gas Company, Docket 50-395, Licensee Event Report 83-110, October 10, 1983.
7.
Letter from O. W. Dixon, Jr., South Carolina Electric and Gas Company, to H. R. Denton, NRC, October 21, 1983.
(1.5) 8.
Duquesne Light Company, Docket 50-334, Licensee Event Report 83-15, June 17, 1983.
(1.6) 9.
Mississippi Power and Light Company, Docket 50-416, Licensee Event Report 83-126, October 20, 1983.
(1.7) 10.
NRC, IE Information Notice 83-69, October 21, 1983.
11.
Florida Power Company, Docket 50-302, Licensee Event Report 83-23, October 3, 1983.
(1.8) 12.
NRC, Preliminary Notification PNO-I-83-100, September 30, 1983.
13.
Boston Edison Company, Docket 50-293, Licensee Event Report 83-48, October 14, 1983.
14.
NRC/ Region I, Inspection Report 50-293/83-19, October 20, 1983.
These referenced documents are available in the NRC Public Document Room at 1717 H Street NW, Washington, D.C. 2055 for inspection and/or copying for a fee.
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2.0 ABSTRACTS OF OTHER NRC OPERATING EXPERIENCE DOCUMENTS 2.1 Abnormal Occurrence Reports (NUREG-0090) Issued in September - October 1983 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 pro-visions 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-0090 series of documents.
Also included in the quarterly reports are updates of some previously reported abnormal occurrences, and summaries of certain events that may be perceived by the public as significant but do not meet the Section 208 abnormal occurrence criteria.
Date Issued Regoy 9/83 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES:
JANUARY - MARCH 1983, NUREG-0090, VOL. 6, NO. 1 During the report period, there were three abnormal occurrences, all occurring at nuclear power plants licensed to operate. The first involved a main feedwater break due to water hammer at Maine Yankee on 1/25/83, the second involved deficiencies in management and procedural controls at Brunswick Units 1 and 2 occurring over a period from 6/28/82 to 7/16/82, and the third involved a failure of the automatic reactor trip system at Salem Unit 1 on 2/22/83.
Also, the report provided update information on the following occurrences previously reported in NUREG-0090:
(1) cracks in pipes at BWRs, first reported in NUREG 75/0090 (January-June 1975); (2) environmental qualification of safety-related electrical equipment inside containment, first reported in NUREG-0090-10 (October-December 1977); (3) the accident at Three Mile Island, first reported in NUREG-0090, Vol. 2, No. 1 (January-March 1979); (4) radiological contamination from well logging operations, first reported in NUREG-0090, Vol. 5, No. 3 (July-September 1982).
In addition, (1) a radioactive release at Browns Ferry on 1/16/83,(2) uranium found in cloisonne jewelry in New York State on 1/25/83, (3) damaged fuel cladding at Farley on 1/15/83, and (4) plant construction deficiencies at Midland Nuclear Power Station were discussed as items of interest that did not meet abnormal occurrence criteria.
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2.2 Bulletins and Information Notices Issued in September - October 1983 The Of fice of Inspection and Enforcement periodically issues bulletins and information notices to licensees and holders of construction permits.
During the period, one bulletin supplement and 14 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 should have to assess the need for further actions.
A prompt response by affected licensees is required and failure to respond appropriately may result in an enforcemnt 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, Institute of Nuclear Power Operations, nuclear steam 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 intenJed as substitutes for revised license condi+. ions or new requirements.
Information Notices are rapid transmittals of information which may not have been completely analyzed by NRC, but which licensees should know.
They require no acknowledgement or response, but recipients are advised to consider the applicability of the information to their facility.
Date Bulletin Issued Subject 83-07, 10/26/83 APPARENTLY FRAUDULENT PRODUCTS SOLD BY Supplement RAY MILLER, INC.
Included as a supplement to 83-07 is IE's report "Apparently Fraudulent Products Sold by Ray Miller, Inc."
The supplement provides information relating to an NRC letter sent to each non-licensee company that received apparently fraudulent material fr:m Ray Miller, Inc., asking the company to identify nuclear facilities that may have been supplied fraudulent material.
The supplement provides a composite list of the respondents and their customers as identified in the responses to the NRC.
Information Date Notice Issued Subject 83-59 9/15/83 DOSE ASSIGNMENT FOR WORKERS IN NON-UNIFORM RADIATION FIELDS All nuclear power reactor facilities holding an operating license or construction permit, research 14
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1 Information Date 3
Notice Issued Subject T.b[a -'
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m and test facilities, fuel cycle facilities, and all j p.
material licensees were provided guidance on proper uC*
dose assignments to workers in non-uniform radiation ik.
fields. This was in response to requests from several
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power reactor licensees regarding proper assignment J
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of extremity and whole body doses to workers involved
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in " sludge lancing" - a specific steam generator
.g'q maintenance activity.
Also included as attachments C-r L
were Information Notice No. 81-26, Part 3, " Place-ment of Personnel Monitoring Devices for External Radiation Exposure," issued 8/28/81, and its Supple-5 ment 1, issued 7/19/82.
1 83-60 9/22/83 FALSIFICATION OF TEST RESULTS FOR PROTECTIVE COATINGS P
All holders of a nuclear power reactor operating license or construction permit and nuclear fuel cycle licensees were provided information pertaining to the alteration of a test report and subsequent 5
use of substandard materials at a nuclear facility.
Earlier this year, a jury found Con-Chem, Inc. (CCI),
a protective coatings company, guilty of fraudulent
.. j practices during 1979 and 1980.
Nuclear power
[
reactor licensees who had purchased coatings from this company during the period of concern were a
notified at the time.
The criminal trial was con-
}
cluded on 3/25/83, and details of the case were i
publicized in this notice to illustrate the risk E
involved in accepting test reports that are not E
provided directly from an independent testing g
facility.
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83-61 9/26/83 ALLEGED USE OF STAND-INS FOR WELDER QUALIFICATION E
TESTS 1
All nuclear power facilities holding an operating Fr license or construction permit were informed that E
the International Brotherhood of Boilermakers Union were selling illegal membership books.
The books indicate that a welder has completed his apprentice-t ship, which enables him to work as a journeyman I
welder.
However, some qualified welders were taking the qualification test on behalf of those who had purchased union membership books but did not have adequate training to pass the test.
Recent NRC a
review of utility practices for establishing welder r
i identity has indicated that in addition to requiring
?
the welder's signature and social security number, some utilities also require a current photograph.
h 15
Information Date Notice Issued Subject The NRC has initiated a request to the ASME to provide a code requirement for positive identifi-cation of the individual taking the welder qualifi-cation test.
83-62 9/26/83 FAILURE OF REDUNDANT T0XIC GAS DETECTORS POSITIONED AT CONTROL ROOM VENTILATION AIR INTAKES All nuclear power reactor facilities holding an operating license were informed that from 1977 through the date of this notice, approximately 68 l
licensee event reports have involved the failure of one or more chlorine and ammonia detectors positioned at the air intakes of control room ventilation systems.
The toxic gas detector most susceptible to failure utilizes a dripping electrolyte. Redundant toxic gas detectors are exposed to the same intake airflow for ventilation systems.
All detectors may fail from a common cause, such as exposure to the l
same source of dirty air.
Personnel at some facili-l ties are considering either increased surveillance or replacement with a more reliable type of detector.
83-63 9/26/83 POTENTIAL FAILURE OF WESTINGHOUSE ELECTRIC CORPORATION TYPE SA-1 DIFFERENTIAL RELAYS All holders of a nuclear power reactor operating license or construction permit were notified of a potentially significant problem pertaining to random trip output of Westinghouse SA-1 differential relays.
A random differential relay trip output will trip the breakers of a differential circuit without the presence of a fault, thereby isolating a transformer, generator, switchgear, or system bus unnecessarily.
Testing by Westinghouse concluded that certain silicon-controlled rectifiers in the relay trip output circuits can cause the random trip output.
The notice provided a list of corrective actions which Westinghouse had sent to its SA-1 relay customers, listed in Attachment 1 to the notice.
However, since most of these customers are not the end-users, it was suggested that licensees review the relay serial numbers in Attachment 1 and compare them with the serial numbers on SA-1 relays in their nuclear facilities to determine if corrective action is required.
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Information Date Notice Issued Subject 83-64 9/29/83 LEAD SHIELDING ATTACHED TO. SAFETY-RELATED SYSTEMS WITHOUT 10 CFR 50.59 EVALUATIONS All nuclear power reactor facilities holding an operating license or construction permit were informed of an event at Maine Yankee (a PWR) where significant quantities of lead shielding were installed on safety-related systems without a proper engineering evaluation as required by 10 CFR 50.59.
Licensees are devoting increased attention and resources to reduce radiation fields in an effort to minimize worker exposure.
Although the NRC encourages these ALARA efforts, this event and other similar occurrences illustrate the need to reemphasize the requirements of 10 CFR 50.59.
It was noted that IE Circular No. 80-18, "10 CFR 50.59 Safety Evaluations for Changes to Radioactive Waste Treatment Systems,"
provides general guidance and clarification regarding the 10 CFR 50.59 requirements.
83-65 10/7/83 SURVEILLANCE OF FLOW IN RTD BYPASS LOOPS USED IN WESTINGHOUSE PLANTS All Westinghouse nuclear power reactor facilities holding an operating license or construction permit were alerted to the possibility of low flow through a reactor coolant loop resistance temperature detector (RTD) bypass tube. An event at Salem on 1/29/83 was cited as an example.
How flow through the RTD bypass line can degrade the response of the temperature measurements used in the reactor protection system.
The presence of low flow in this line should be revealed by a low flow alarm. Although no regulatory requirements or vendor recommendations are estab-lished regarding periodic surveillance of the flow sensor or alarm, Westinghouse licensees may find it prudent to calibrate the-flow sensors on a refueling outage basis and to verify the alarm setpoint on a monthly basis to assure the operability of this monitoring function.
83-66 10/7/83 FATALITY AT ARGENTINE CRITICAL FACILITY All nuclear power reactor facilities holding an operating license or construction permit, and non-power, critical facility, and fuel cycle licensees were-informed of an accident involving a fatality at a zero power critical test and training facility owned and operated by the Argentine National Atomic 17
Information Date Notice Issued Subject Energy Commission.
With the reactor sub-critical (shut-down), an operator was making core configura-tion changes using an overhead crane.
Facility procedures required that fuel and control rod alterations be performed without the moderator present.
The qualified operator (14 years of experience) attempted to make core changes without draining the moderator water.
The core went prompt critical (estimated integrated energy pulse approxi-mately 10 megajoules). The moderator expanded rapidly, shutting down the reactor, followed by an automatic dump of the moderator.
There was no equipment damage or significant radiation exposure to personnel other than the operator.
83-67 10/11/83 EMERGENCY-USE RESPIRATOR MATERIAL DEFECT CAUSES PRODUCTION OF N0XIOUS GASES All nuclear power reactor facilities holding an operating license or construction permit, research and test reactor licensees, fuel cycie licensees, and Priority I material licensees were informed of a potentially serious problem with the Bio-Pak 60-P respirator manufactured by Rexnord Company (NIOSH/MSHA approval number TC 13F-85).
This respirator is approved for emergency use by the National Institute for Occupational Safety and Health (NIOSH) and is a closed circuit, positive pressure self-contained breathing apparatus (SCBA).
The oxygen supply valve seat of the high pressure oxygen bottle (manufactured prior to 1981) tends to shear during valve operation, l
creating Kel-F(TM) fibers.
During startup for opera-tion at San Onofre in July 1983 these fibers ignited in the pure oxygen supply stream, releasing combus-tion products including C0, CO, CF and HF.
It is 2
4 expected that licensees will review the information for applicability to their facilities.
Further NRC action may result from feedback from the ongoing NIOSH review effort.
83-68 10/11/83 RESPIRATOR USER WARNING:
DEFECTIVE SELF-CONTAINED BREATHING APPARATUS AIR CYLINDERS All nuclear power reactor facilities holding an operating license or construction permit, research and test reactors, fuel cycle licensees, and Priority I material licensees were informed of a serious, potential defect in certain self-contained breathing apparatus (SCBA) fiberglass wound aluminium 18
i Information Date Notice Issued Subject cylinders which contain high pressure (4500 psi) air.
Should a cylinder fracture, serious injury or death and/or property damage could result.
In two cases, the defective cylinders ruptured but no injuries were reported.
The National Institute for Occupational Safety and Health (NIOSH) issued a Respirator Users Warning, provided as Attachment 1.
Also, included as Attachment 2 is NIOSH's " Guide-lines for Reporting Respirator Problems." At the manufacturers request, the product recall notice was issued by the Department of Transportation's Office of Hazardous Material Regulation on 8/11/83.
83-69 10/21/83 IMPROPERLY INSTALLED FIRE DAMPERS AT NUCLEAR POWER PLANTS All nuclear power reactor facilities holding an operating license or construction permit were informed of three related potentially generic problems involving the improper installation of fire dampers in ventilation ducts which penetrate fire barriers in safety-related areas.
Any one, or all three problems may exist at a facility.
These dampers were installed within ti.e ducts in the fire area and not within the fire wall penetration as required by the design and installation code, National Fire Protection Association Standard 90A, Air Conditioning and Ventilatirig Systems.
(See also pp. 9-10.)
83-70 10/25/83 VIBRATION INDUCED VALVE FAILURES All nuclear reactor facilities holding an operating license or construction permit were notified of several events at nuclear power plants in which valve failures and system inoperability were the result of normal operational vibration.
Discussions with General Electric indicate that some of these failures may be generic to a specific valve type.
The notice discussed events at Quad Cities and Browns Ferry involving loosened studs, nuts, and/or bolts on a recirculation pump discharge valve, a suppression pool return valve, and a low pres-sure coolant injection valve.
In addition, problems with loose or missing stem clamp setscrews on Anchor Darling globe valves at Shoreham and Zimmer were noted.
At Shoreham, the failure of a resi-dual heat removal globe valve to stroke during testing was especially significant in that the 19 l
Information Date Notice Issued Subject remote valve position indication was from limit switches on the motor operator, so that the valve appeared to be opening and closing normally when in fact the valve had not moved.
The notice stressed that station personnel should be aware of the potential for vibrational loosening of valve components and may want to emphasize this aspect in valve preventive maintenance.
83-71 10/27/83 DEFECTS IN LOAD-BEARING WELDS ON LIFTING DEVICES FOR VESSEL HEAD AND INTERNALS All holders of nuclear power reactor operating licenses or construction permits were informed of the potential for failure of reactor head and internal lifting devices manufactured by Babcock and Wilcox Company due to defective load-bearing welds.
If defective load-bearing welds were to go uncor-rected, they could jeopardize the capability of the lifting system to safely handle safety-related components within reactor compartments.
The subject lifting device was fabricated to industry standards.
However, a subsequent regulatory requirement, NUREG-0612, requires that the lifting device be tested in accordance with ANSI N14.6 on the basis that it is to carry heavy loads over safety-related systems.
83-72 10/28/83 ENVIRONMENTAL QUALIFICATION TESTING EXPERIENCE All holders of a nuclear power reactor operating license or construction permit were informed of environmental qualification test failures.
These test failures are based on (1) Construction Deficiency Reports and 10 CFR Part 21 Reports sub-mitted to the NRC, and (2) results from the NRC-sponsored environmental qualification methodology research program.
The notice also serves to inform the licensees of findings that resulted from inspec-tions conducted by the licensee or its agent of equipment that has been environmentally qualified and is being delivered or installed at the sites.
83-73 10/31/83 RADIATION EXPOSURE FROM GLOVES CONTAMINATED WITH URANIUM DAUGHTER PRODUCTS All licensees authorized to process uranium as source taaterial and metal producers of alloys 20
Information Date Notice Issued Subject containing uranium as source material except uranium mills. UF-6 facilities, uranium fuel fabrication plants and nuclear power plants were advised to bring to the attention of all persons involved in the administration and operation of uranium processing facilities a recent incident that resulted in multiple exposures to workers' hands in excess of regulatory limits, and to discuss the generic implications for other, similar operations.
One of the licensee's health physics technicians found that the leather gloves worn by the foundry workers were routinely contaminated during handling of the bricks and crucible upon completion of a melt. Measurements made with a Juno-type survey meter, with the gloves turned inside out, showed radiation levels inside the gloves between 270 and 1000 millirems per hour. While foundry workers had been provided with wrist badges, these wrist badges did not adequately measure the exposure to workers' hands.
The badges were located too far from the hands that were receiving most of the radiation exposure.
An investigation was conducted to determine the workers' hand exposures during the fourth quarter of 1982 and the first quarter of 1983.
It was estimated from TLD measurements and interviews with foundry workers that beta radiation dose rates of abaut 1 rem per hour existed inside the contaminated gloves.
Based on those estimated dose rates, from 10 to 15 foundry workers received hand exposurec of up to 125 rems per quarter during both quarters.
Similar doses may have been receivrd in previous quarters.
The licensee's investigation is continuing.
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2.3 Case Studies and Engineering Evaluations Issued in September - October 1983 The Office for Analysis and Evaluation of Operational Data (AE00) has as a primary responsibility the task of reviewing 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 the effort involved usually is a few staffweeks of investigative time.
Case studies are in-depth investigations of apparently 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 ard prevent recurrence of the event.
Before issuance, this report is sent for peer review and comment to at least the applicable utility and appropriate NRC offices.
These AEOD 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 Subject NUREG/CR-3122 9/83 POTENTIALLY DAMAGING FAILURE MODES OF HIGH-AND ORNL/NSIC-213 MEDIUM-VOLTAGE ELECTRICAL EQUIPMENT A study was conducted for AE0D by Oak Ridge National Laboratory on operating experience involving high-and medium voltage electrical equipment.
The report (NUREG/CR-3122) considered the electrical faults of transformers, switchgear (circuit breakers), lighting arrestors, high voltage cabling and buses, and other electrical equipment which through failure can be the initiating event that may expand the original fault to nearby or associated equipment.
Recommendations from the study included:
(1) those of a general nature that apply to the entire electrical system and involve such activities as better quality assurance, better procedures, better failure documentation, and better information exchange; and (2) those specific to individual electrical components.
The recommendations remain under study to determine if further regulatory action is warranted.
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Engineering Date Evaluation Issued Subject E319 9/8/83 LOSS OF DRYWELL-TORUS PRESSURE DIFFERENTIAL DURING RESIDUAL HEAT REMOVAL PUMP FLOW TESTING AT COOPER NUCLEAR STATION A licensee event report submitted for Cooper Nuclear Station described an event in which one or more of the primary containment downcomers were uncovered as a result of suppression pool surface waves caused by residual heat removal (RHR) system pump flow testing.
An evaluation of this event revealed that the poten-tial for suppression pool waves caused by RHR system pool cooling flows had not been considered in the generic downcomer uncovery analysis for the Mark I containment system with reduced downcomer submergence.
A supplementary analysis performed for this evalua-tion shows that downcomer uncovery might occur if peak RHR flow-induced pool waves were added to the previously considered pool surface level reduction e f fects.
However, further assessment shows that even if limited downcomer uncovery were to occur during pool cooling after a loss-of-coolant acci-dent, the calculated peak suppression pool pressure would not increase significantly above that which previously had been predicted.
E320 9/8/83 POWER OPERATED RELIEF VALVE (PORV) ACTUATION RESULTING IN SAFETY INJECTION ACTUATION On February 3, 1983 at 6:03 p.m., with Calvert Cliffs Unit 2 in Mode 3 (hot standby), both power operated relief valves (PORVs) opened as a result of an operator mistakenly deenergizing reactor protec-tive system (RPS) channel D, prior to a test of the No. 21 vital 120 V ac bus power transfer switch. As a result of the occurrence, the reactor coolant system pressure dropped from 2250 psia to 1520 psia before the operators diagnosed the event and closed the PORVs.
The PORVs were open for about 30 seconds.
The mass and energy release through the PORVs was sufficient to blow out the rupture disk in the quench tank. As expected, a partial safety injection actua-
. tion (channel B) occurred when the reactor coolant system (RCS) pressure decreased below the 1780 psia' -
safety injection actuation signal setpoint. Since the RCS pressure never dropped below the 1260 psia shutoff head of the high pressure safety injection pumps, no. water was actually injected into the RCS.
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Engineering Date Evaluation Issued Subject A review of the PORV actuation logic and the control circuitry reveal an inconsistency in the failure positions of the PORVs on loss of electric power.
If, for example, the 480 V ac power source to a PORV's solenoid was deenergized, the PORV would fail closed.
If, however, two or more high pressure bistable trip units were deenergized, both PORVs would fail open.
This inconsistency is because the output of these bistables is not only input to the RPS, causing the reactor trip on high system pres-sure, but is also input to the PORV actuation con-trol logic to open the PORVs.
Since the high pres-sure bistable trip units provide a reactor trip function, their failure mode is in the tripped posi-tion on loss of power.
While the tripped position is the fail-safe position for the RPS, since it causes a reactor shutdown, it will also cause the PORVs to fail open, creating a reactor blowdown.
It is recognized that two independent failures, i.e., loss of two 120 V ac vital instrument buses, are required to cause the PORVs to fail open.
- Also, hand switches are provided on the control board allowing plant operators to either close the PORVs, by overriding the "open signal" from the RPS, or to close the block valves.
Therefore, an uncontrolled blowdown of the RCS due to the PORVs failing open is unlikely.
The evaluation suggested that the PORV actuation logic at plants where Combustion Engineering is the nuclear steam system supplier be reviewed and, if necessary, changed to allcw the PORVs to fail closed on the deenergization of two or more RPS channels.
E321 9/12/83 THREE SIMILAR EVENTS OF A LOSS OF SHUTDOWN COOLING FLOW AT CE PLANTS Three short interruptions of shutdown cooling (SDC) flow occurred at two different Combustion Engineering (CE) units in a relatively short period of time.
The initiating causes were similar, a disruption in electrical power.
In each case, the pressure indi-cator-controller (PIC) failed high on loss of power, causing the pressure interlocks to isolate the suction valves of the SDC system in response to the perceived high reactor coolant system pressure.
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Engineering Date Evaluation Issued Subject The evaluation concludes that the loss of SDC flow in the three events was not'an important safety concern.
However, the two interruptions of SDC flow during refueling may have been unnecessary because the reactor coolant system (RCS) cannot be pres-surized with the reactor vessel head removed.
In cold shutdown, with the RCS intact, overpressure protection is provided by a relief valve in the suction line of the SDC system.
During refueling, the overall reliability of the SDC system may be improved by temporarily bypassing the PIC interlocks.
E322 9/16/83 DAMAGE TO VACUUM BREAKER VALVES AS A RESULT OF RELIEF VALVE LIFTING Licensee Event Report 82-036 for Peach Bottom 2 describes two events in which the vacuum breaker valves on a safety relief valve tail pipe were found damaged after actuation of safety relief valves.
Two additional units found in this review have had similar occurrences. One was Browns Ferry I and the other was Hatch 2. There were five vacuum breaker valves found to be defective in each-unit, with a total of 12 damaged vacuum breaker valves in the three events.
Although the valves in these three units are dif-ferent in size and brand, the damage to them seems to be very similar, and redesign of some internal parts has been included in the repair of all these valves.
Therefore, design deficiency could con-tribute to the damage. This also implies that the qualification program may be inadequate. The failure of a vacuum breaker valve may cause steam discharge to the drywell, and subsequently create a small LOCA environment in the drywell. There is a possibility that a subsequent actuation of the SRV associated with the damaged vacuum breaker valve could lead to a large hydrodynamic load on the piping and torus beyond the design condition.
E323 9/18/83 LOAD REDUCTION TRANSIENT AT THE SALEM NUCLEAR POWER PLANT, UNIT 2 ON JANUARY 14, 1982 This study evaluates the significance and safety implications of the January 14, 1982 load reduction transient at Salem Unit 2.
The event involved five separate and unrelated failures including a complete 25
Engineering Date Evaluation Issued Subject failure of the rod control system, and was initiated by a feedwater transient involving a loss of suction to the feedwater pumps.
A power mismatch of 70*;
between reactor power and turbine load occurred during the event.
Elevated average temperature (Tavg) exceeded the technical specification limits, but plant stability was restored without a trip to preclude an overcooling transient. Although the scram function was available, there was no require-ment for a reactor trip when the rod control system had failed and the Tavg had exceeded technical specification limits.
The AEOD evaluation of the Salem event found that the most significant aspect was the failure of the rod control system following a load reduction transient, which resulted in an elevated average temperature exceeding the technical specifications by 10 F.
The other failures during the event were found to be unrelated, randomly occurring failures.
This study also evaluates the causes for the i
transient, system response, operator actions, and I
the generic implications of this and similar events.
E324 9/21/83 REVIEW OF EVENTS INVOLVING FAILURES OF POWER SUPPLY IN INSTRUMENTATION AND CONTROL SYSTEM Licensee Event Reports (LERs) in the Sequence Coding and Search System (SCSS) data base were searched for events involving failures of power supply units in Instrumentation and Control (I&C) Systems at oper-ating nuclear plants.
(Events involving inverters, battery chargers and instrument buses were excluded).
From this data, a total of 86 events involving power supply failure were obtained.
The relevant LERs were tabulated and reviewed.
The causes of failure, when stated in the data, were evaluated.
Patterns of failure were obtained, depending on the type of I&C system in which power supply failure occurred.
Certain trends of repetitive failure at particular plants were found.
The two patterns of failures that predominated are in (1) the radiation monitoring instrumentation system, and (2) the excore nuclear instrumentation system.
However, since the LERs did not provide sufficient detailed information regarding the make, model and manufacturer of the failed power supply units, the reviewers were unable to come to any definite conclusions regarding the patterns found.
A more detailed trends and patterns study on the same subject is planned, to obtain more refined 26
Engineering Date Evaluation Issued Subject findings and conclusions.
In this study, trends were derived from repetitive failures at five plants.
Licensees at two of the plants are already aware of these trends and are taking adequate corrective actions to address them. At the other three plants the licensees apparently have not formulated any long term corrective action as yet.
Short term corrective actions were taken at all plants where failures have occurred.
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2.4 Generic Letters Issued in September-October 1983 Generic letters are issued by the Office of Nuclear Reactor Regulation, Division of Licensing.
They are similar to IE Bulletins (see Section 2.2) in that they transmit information to, and obtain information from, reactor licensees, appli-cants, and/or equipment suppliers regarding natters of safety, safeguards, or environmental significance.
During September and October 1983, three letters were issued.*
Generic letters usually either (1) provide irformation thought to be important in assuring continued safe operation of facilities, or (2) request information on a specific schedule that would enable regulatory decisions to be made regard-ing the continued safe operation of facilities.
They have been a significant means of communicating with licensees on a number of important issues, the resolutions of which have contributed to improved quality of design and operation.
Generic Date Letter Issued Subject l
83-31 9/19/83 SAFETY EVALUATION OF " ABNORMAL TRANSIENT OPERATING GUIDELINES" The NRC staff has reviewed the proposed Oconee Nuclear Station, Unit 3 Abnormal Transient Operating Guidelines (ATOG) as described in Babcock & Wilcox (B&W) Owners Group letters dated March 31, 1982 and June 15, 1982, and D.
Napior's letter from B&W to the Owners Group dated March 14, 1983. As discussed in the enclosed letter to the B&W Owners Group, the NRC has concluded that ATOG is acceptable as a basis for implementation of improved plant-specific proce-dures and will provide improved guidance for operator emer-gency procedures over that which currently exists, Since there is no generic version of ATOG for B&W plants, the utilities who are participating in the Owners Group program are to provide sufficient documentation in the form of plant-specific AT0Gs and Transient Information Documents (TIDs) so that the NRC can perform comparisons with the ATOG version evaluated in the enclosed Safety Evaluation Report (SER).
Suggested implementation of the guidelines was pro-vided to all operating reactor licensees, applicants for an operating license, and holders of constructior, permits for B&W pressurized water reactors.
- Generic Letter 83-34 has not yet been issued; Generic Letters 83-32, 83-35, 83-36, and 83-37 will be abstracted in Power Reactor Events Vol. 5, No. 6, covering documents issued during November-December 1983.
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t Generic Date Letter Issued Subject 83-33 10/19/83 NRC POSITIONS ON CERTAIN REQUIREMENTS OF APPENDIX R TO 10 CFR 50 During evaluations of exemption requests, the NRC determined that some licensees were interpreting certain requirements of Appendix R in a manner that was not consistent with the position that the staff was using. Where any such differ-ences were discovered, NRC informed these licensees in the NRC Safety Evaluation Report supporting the granting or denial of an exemption.
More recently, inspections for con-formance to Appendix R were completed at four plants, the licensees for which had indicated that all modifications for conformance had been completed or other modifications ap-proved by exemptions had been completed.
In these inspec-tions, the NRC inspection team also identified what the staf f considers to be nonconformance with requirements of Append;x R, for which exemptions had not been requested or justified.
Therefore, this generic letter transmitted an enclosure on NRC staff positions on requirements of Appen-dix R to all licensees and applicants of nuclear power reactors for information and use as appropriate.
The NRC inspection teams that will be conducting inspections for conformance to Appendix R at each plant will be using these positions as their criteria for conformance for these particular issues.
83-38 10/31/83 NUREG-0965, "NRC INVENTORY OF DAMS" All licensees of operating reactors and applicants for oper-ating licenses were provided a copy of the subject report, which describes in detail the NRC Dam Safety Program.
The information provided in this report was obtained from NRC files such as safety analysis reports.
The primary purpose of the report is to identify all dams that could, in any way, be considered "NRC dams" in the context of the Guide-lines of the Federal Dam Safety Program. A secondary pur-pose is to identify the dams for which the Federal Guide-lines may be applicable and for which the NRC might be considered " responsible."
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2.5 Operating Reactor Event Memoranda Issued in September-October 1983 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 memorandum (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 g
Operational Data, and of Inspection and Enforcement for their information.
No OREMs were issued during September-October 1983.
m
t 2.6 Regulatory and Technical Reports (NUREG-0304) Issued In September -
October 1983 The abstracts listed below have been selected from the Office of Administration's quarterly publication, Regulatory and Technical Reports (NUREG-0304).
This document compiles abstracts of the formal regulatory and technical reports issued by the NRC staff and its contractors.
Bibliographic data for the reports are also included.
Copies and subscriptions of NUREG-0304 are available from the NRC/GPO Sales Program, PHIL-016, Washington, DC 20555 or on (301) 492-9530.
Report Title NUREG-0020 LICENSED OPERATING STATUS
SUMMARY
REPORT Vol. 7, No. 6 (Vol. 7, No. 6 - data as of 5/31/83; Vol. 7, No. 7-September 1983; data as of 6/30/83)
Vol. 7, No. 7 October 1983 This report provides data on the operation of nuclear units as timely and accurately as possible.
This infor-mation is collected by the Office of Resource Management from the Headquarters staff of NRC's Office of Inspection and Enforcement, from NRC's Regional Offices, and from utilities.
The three sections of the report are:
monthly highlights and statistics for commercial operating units, and errata from previously reported data; a compilation of detailed information on each unit, provided by NRC's Regional Offices, IE Headquarters, and the utilities; and an appendix for miscellaneous information such as spent fuel storage capability, reactor years of experience and non power reactors in the U.S.
It is hoped the report is helpful to all agencies and individuals interested in maintaining an awareness of the U.S. energy situation as a whole.
NUREG-0900 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES Vol. 6, No. 1 (January - March 1983)
September 1983 Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled inci-dent or event which the NRC determines to be significant from the standpoint of public health or safety and requires a quarterly report of such events to be made to Congress.
This report covers the period 1/1/83 through 5/31/83.
During the report period, there were three abnormal occurrences at the nuclear power plants licensed by the NRC to operate.
The first involved a main feedwater line break due to water hammer.
The second involved management 4
31
Report Title and procedural control deficiencies.
The third involved failure of the automatic reactor trip system.
There were no abnormal occurrences 'or the other NRC licensees.
There were six abnormal occurrences at Agreement State licensees.
One involved an individual who ingested and was contaminated by radioactive material.
Four involved lost or stolen radioactive sources.
One involved radio-active contamination of a metals production facility.
NUREG-0540 TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE (Vol. 5, Vol. 5, No. 7 No. 7 - July 1983; Vol. 5, No. 8 - August 1983)
September 1983; Vol. 5, No. 8 This document is a monthly publication containing October 1983 descriptions of information received and generated by the NRC.
This information includes (1) docketed material associated with civilian nuclear power plants and other uses of radioactive materials, and (2) nondocketed material received and generated by NRC pertinent to its role as a regulatory agency.
The following indexes are included:
Personal Author Index, Corporate Source Index, Report Number Index, and Cross Reference to Principal Documents Index.
NUREG-0580 REGULATORY LICENSING STATUS
SUMMARY
REPORT (Vol. 12, Vol. 12, No. 8 No. 8 - August 1983; Vol. 12, No. 9 - September 1983)
September 1983; Vol. 12, No. 9 This report provides a review of the status of the progress October 1983 of the licensing reviews for all construction permits, operat-ing licenses, special projects and non power reactor renewals under review, as reported to Congress.
NUREG-0606 UNRESOLVED SAFETY ISSUES
SUMMARY
(August 1983)
Vol. 5, No. 3 September 1983 This report provides an overview of the status of the progress and plans for resolution of the generic tasks addressing " Unresolved Safety Issues" as reported to Cungress.
NUREG-0714 OCCUPATIONAL RADIATION EXPOSURE REPORT (Vol. 2 - 1980; Vols. 2 and 3 Vol. 3 - 1981)
October 1983 These reports summarize the information reported for calendar years 1980 and 1981 by all NRC licensees to the Commission's centralized repository of personnel occupa-tional radiation exposure information.
The bulk of the information in the report is derived from annual reports that were required to be submitted by all NRC licensees pursuant to 10 CFR 20.407.
All NRC licensees were required to submit an annual exposure report.
The requirement of 10 CFR 20.408 for the submission of termination reports continued to apply to commercial nuclear power reactors, 32
Report Title industrial radiographers, fuel fabricators and processors and commercial distributors of byproduct material, and some analysis of the data contained in these reports is presented.
A brief description of personnel overexposures reported by NRC licensees is included also.
NUREG-1021 OPERATOR LICENSING EXAMINER STANDARDS October 1983 This report provides policy and guidance to NRC examiners and establishes the procedures and practices for examining and licensing of applicants for NRC operator licenses pursuant to Part 55 of Title 10 of the Code of Federal Regulations (10 CFR 55).
It is intended to assist NRC Examiners and facility licensees to understand the exam-ination process better and to provide for equitable and consistent administration of examinations to all appli-cants by NRC examiners.
This standard is not a substitute for the Operator Licensing Regulations.
As appropriate, this standard will be periodically revised to accommodate comments and reflect new information or experience.
NUREG-1022 LICENSEE EVENT REPORT SYSTEM September 1983 On July 26, 1983, the Commission published in the Federal Register a final rule (10 CFR 50.73) that modifies h
and codifies the Licensee Event Report (LER) system.
The rule became effective on January 1, 1984.
This NUREG provides supporting information and guidance that will be of interest to persons responsible for the preparation and review of LERs.
It also is useful for guidance on the revised 10 CFR 50.72 criteria (immediate notification of significant events) that became effective on January 1, 1984.
The information contained in this NUREG includes:
(1) a brief description of how LERs are analyzed by the NRC, (2) a restatement of the guidance contained in the Statement of Consideration that accompanied the publica-tion of the LER rule, (3) a set of examples of potentially reportable events with staff comments on the actual report-ability of each event, (4) guidance on how to prepare an LER, including the LER forms, and (5) guidance on submittal of LERs.
NUREG/CP-0047 TRANSACTIONS OF THE ELEVENTH WATER REACTOR SAFETY RESEARCH September 1983 INFORMATION MEETING This report contains summaries of papers on reactor safety research work presented at the lith Water Reactor Safety Research Information Meeting.
The meeting was held at the National Bureau of Standards in Gaithersburg, Maryland in October 1983.
The summary reports highlight the programs 33
Report Title and results of nuclear safety research work sponsored by the Office of Nuclear Regulatory Research, NRC.
Summaries of invited papers are also included.
The latter represent work on reactor safety research conducted by the electric utilities through the Electric Power Research Institute, the nuclear industry, and various government and industry I
organizations in Europe and Japan.
3-4 NUREG/CR-2000 LICENSEE EVENT REPORT COMPILATION (Vol. 2, No. 8 -
h Vol. 2, No. 8 August 1983; Vol. 2, No. 9 - September 1983)
September 1983; h
Vol. 2, No. 9 This monthly report contains Licensee Event Report (LER)
October 1983 operational information that was processed into the LER data file of the Nuclear Operations Analysis Center (NOAC) during the one month period identified on the cover of this document.
The LERs, from which this information is derived, are submitted to the NRC by nuclear power plant licensees in accordance with Federal regulations.
Procedures for LER reporting are described in detail in the NRC Regula-tory Guide 1.16 and NUREG-0161, Instructions for Prepar-ation of Data Entry Sheets for Licensee Event Reports.
The LER summaries in this report are arranged alphabet-ically by facility name, system, and keyword indexes follow the summaries.
The components and systems are those identified by the utility when the LER form is initiated; the keywords are assigned by the computer using
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correlation tables from the Sequence Coding and Search System.
NUREG/CR-2238 ADVANCED REACTOR SAFETY RESEARCH QUARTERLY REPORT Vol. 4 (10/81 through 12/81)
September 1983 Sandia National Laboratories is conducting, under NRC's spsonsorship, phenomenological research related to the safety of commercial nuclear power reactors.
The overall objective of this work is to provide NRC a comprehensive data base essential to (1) defining key safety issues, (2) understanding risk-significant accident sequences; (3) developing and verifying models used in safety assess-ments, and (4) assuring the public that power reactor systems will not be licensed and placed in commercial service in the U.S. without appropriate consideration being given to their effects on health and safety.
This report describes progress in a number of activities dealing with current safety issues relevant to both light water and breeder reactors.
The work includes a broad range of experiments to simulate accidental conditions to provide the required data base to understand important accident sequences and to serve as a basis for development and r.
34
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Title verification of the complex computer simulation models and codes used in accident analysis and licensing reviews.
NUREG /CR-3110 RELIABILITY'0F NONDESTRUCTIVE EXAMINATION L
Vols. I through III This 18-chapter, three-volume study evaluates the various October 1983 nondestructive examination (NDE) techniques now used to detect flaws in components of nuclear systems so that'the reliability of the techniques may be increased.
The signi-ficance of flaws at various. locations in pressure boundary components are assessed along with ways to optimize the NDE procedures needed to detect, locate and size them. _ Emphasis is placed on an integrated program which also considers design, fabrication procedures, and materials.
The data available on the reliability-of' detecting, locating and sizing flaws by NDE are used to construct a probabilistic fracture mechanics model.
The model highlights the signi-ficance of the failure to detect flaws, and to accurately locate or size them in the context of corrponent failure probability.
The study was conducted under-ti e NRC program on;the
" Integration of NDE Reliability and Fracture Mechanics."
Its objectives include (1) improving examination procedures for incorporation into the American Society for Mechanical h.
Engineers (ASME), Boiler and Pressure Vessel Codes, Sec-tion III, V, XI; and (2) gaining a better insight into the influence of irtproved reliability of NDE(in detecting, locating and sizing flaws on component failure probabilities.
NUREG/CR-3177 METHODS FOR REVIEW AND EVALUATION OF EMERGENCY PROCEDURE Vol. 3' GUIDELINES October.1983 Systematic methods for finalizing, reviewing, or developing improved emergency procedure guidelines are applied to a representative General Electric BWR plant.
The methods are based on thr, use.. of operator action event trees (OAETS)'
which document theckey operator actions and plant symptoms associated with the various; stages of risk significant multiple failure accident-sequences.
The application of~
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the methodology utilizes OAETS developed for a BWR-4 and
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Mark I containment and -the BWR Owners' Group Emergency Procedure Guidelines (Revision 2).
Those aspects of General Electric plant design, operation, or response to I.
multiple failure-accident sequences which could result in incomplete, ambiguous, or incorrect guidance to the oper-
' r' ator-if nut carefully addressed in the plant-specific guideline developmer.t.;or utilization process ard 4
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i Report Title NUREG/CR-3371 TASK ANALYSIS OF NUCLEAR POWER PLANT CONTROL ROOM CREWS Vols. I and 2 September 1983 A task analysis of nuclear power plant control room crews was performed by General Physics Corporation and BioTech-nology, Inc. for the Office of Nuclear Regulatory Research, US NRC.
The task analysis methodology used in the project is discussed and compared to traditional task analysis and job analysis methods.
The objectve of the project was to conduct a crew task analysis that would provide data for evaluating six areas:
(1) human engineering design of control rooms, (2) the numbers and types of control room operators needed with requisite skills and knowledges, (3) operator qualification and training requirements, (4) normal, off-normal, and emergency operating proce-dures, (5) job performance aids, and (6) communications.
A generic structural framework for assembling the large task data base was employed from observations and video-taping of crew behaviors during 44 operating sequences conducted at eight power plant sites.
The results of the data collection effort was compiled in a computerized task database.
Six demonstrations for verifying the suitability of the analytical approach and for suitability analysis of each of the six areas were performed and described.
Volume 1 details the Project Approach and Methodology, and Volume 2 provides the Data Results including a description of the computerized task analysis data format.
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