ML20083C265
| ML20083C265 | |
| Person / Time | |
|---|---|
| Issue date: | 12/31/1983 |
| From: | Massaro S NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD) |
| To: | |
| References | |
| NUREG-BR-0051, NUREG-BR-0051-V05-N3, NUREG-BR-51, NUREG-BR-51-V5-N3, NUDOCS 8312220246 | |
| Download: ML20083C265 (43) | |
Text
/9) POWER REACTOR EVENTS
/
United States Nuclear Regulatory Commission May-June 1983/Vol. 5, No. 3 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 from 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 at 1717 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/CPO sales Program. (301) 492-9530, or at PHIL-016, Washington. DC 20555.
Table of Contents Page 1.0 SUMMARIES OF EVENTS 1.1 Airborne Radioactivity in Primary Containment...................................
1 1.2 Degraded High Pressure Safety injection System.....................................
3 1.3 Service Water Bay Leak Causes Loss of Flow to Boron injection, Residual Heat Removal, and Diesel Generator S 5
1.4 Failure of Service Water System Valves...............ystems..........................
7 1.5 Reactor Vessel Internals Bolting F_ailures...............................................
8 1.6 Events Involving Two or More Simultaneous Dropped Rod ControI Cluster A ssemblies...........................................
11 1.7 Damage to Thermai Shield...............................................................
14 1.8 References...............................................................................
16 2.0 A BSTRA CTS OF OTHER NRC OPERA TING EXPERIENCE DOCUMENTS 2.1 Abnormal Occurrence Repo'rts (NUREG-0090).....................................
17 2.2 ButIetins and informa tion No tices.......................................................
18 2.3 Engineering Evaluations and Case Studies.........................
27 2.4 Gen eric L e tters...............................................................................
30 2.5 Operating Reactor Event Memoranda....................................................
33 2.6 Regulatory and Technical Reports (NUREG-0304).................................
34 Editor: Sheryl A. Massaro Associate Editor: Steven E. Trenery Office for Analysis and Evaluation of Operational Data U. S. Nuclear Regulatory Commission Published in:
December 1983 Washington, D. C. 20555 8312220246 831231 PDR NUREC GR-0051 R PDR
1.0 SUMMARIES OF EVENTS 1.1 Airborne Radioactivity in Primary Containment At about 5:00 p.m. on April 16, 1983 at Salem Unit 2,* two workers that had exited the radiological control point from containment were found to have low level nasal area contamination.
Subsequent whole body counting showed minor intakes of cobalt-58 and -60.
The workers had been repairing a broken stud on the cold leg nozzle dam of the No. 22 steam generator.
Other personnel who had been in the vicinity of the No. 22 reactor coolant pump and steam generator were found to have facial contamination.
An NRC investigation of the event identified violations involving (1) failure to use an airborne radioactivity removal system, which resulted in personnel contamination, and (2) failure to perform airborne radioactivity measurements for purposes of" determining compliance with 10 CFR 20.103.
The repair work began at about 11:00 a.m. on April 16 with mechanical loop dams (i.e., bolted plates) in the hot and cold legs of the primary loop to permit work in the steam generator to proceed with water in the primary lines during refueling.
The crew on the No. 22 steam generator work platform consisted of three personnel from Westinghouse (the nuclear steam supply system vendor), one contractor tool control individual, and one radiation protection technician.
The individuals entering the steam generator tent wore air-supplied full face respiratory protective equipment.
Those individuals on the platform but not entering the tent wore full face filter respiratory protective equipment.
Prior to entry into the steam generator primary manways (waterbox), the radiation protection technician collected an eight-minute grab air sample inside the waterbox by use of a two cubic foot per minute (CFM) air sarrple pump.
The air sample was analyzed at about 3:00 p.m. on April 16.
A steam generator jumper time keeper was positioned above the No. 22 steam generator to keep a record of each individual's time in the generator. A second radiation protection technician was positioned at the bottom of the ladder leading to the steam generator platform for purposes of assisting individuals in donning additional protective clothing and equipment.
These two individuals did not wear respiratory protective equipment.
At about 1:00 p.m. on April 16, a Wutinghouse worker entered the waterbox to set up lighting and drill out the broken stud. A second Westinghouse worker entered after the first individual's exit to verify drill alignment.
The drilling was performed from outside the waterbox tent.
The exhaust air from the air-powered drill exhausted into the waterbox.
The first segment of drilling was stopped at about 1:30 p.m.
A worker entered the waterbox to blow out shavings from the hole with an air gun and install a second, larger drill bit.
No airborne radioactivity samples were collected in the waterbox during this entry.
The drilling proceeded for about 10 minutes and was again performed from outside the waterbox tent.
Salem Unit 2 is a 1106 MWe (net) PWR located in New Jersey, 20 miles south of Wilmington, Delaware, and is operated by Public Service Electric and Gas.
J At about 1:45 p.m., a worker entered the waterbox to blow out shavings from the hole with an air gun and to install a third, larger dril bit.
During drilling with the third bit, problems were encountered.
It was believed the bit was broken. At about 2:00 p.m., the third drill was backed out and an attempt was made to tap the broken stud.
While in the generator attempting to tap the stud, the individual's stay time elapsed, and he exited the generator.
The health physics crew for the job was changed between 2:00 and 2:30 p.m.
The tapping proceeded, and the No. 22 reactor coolant pump motor, which was decoupled from the pump shaft, was started at about 2:30 p.m. for vibration testing.
While the motor was running, personnel completed the tapping of the stud and cleaned out the hole by use of the air gun.
The air gun was also used to blow out the grooves and bolt holes in the nozzle dam.
The waterbox was then wiped out with a rag and inspected.
The No. 22 reactor coolant pump motor was shut off at about 4:00 p.m., when testing was completed.
At about 5:00 p.m., a technician performing whole body counting of an individual who had worked in the area identified cobalt-58 intakes. A second individual was counted and also found to have an intake of cobalt-58.
An investigation was initiated at about 5:15 p.m.
Air samples collected in the vicinity were qualitatively measured with a thin window GM detector system and were found to indicate high airborne radioactivity. An air sample collected in the vicinity of the No. 22 reactor coolant pump during the period 5:05 to 5:10 p.m. measured 40 millirad / hour with an end window ion chamber survey meter.
The entire containment was evacuated at about 6:15 p.m.
The licensee identified 208 individuals who were in containment during the period. About 150 had been given a whole body count as of 7:00 a.m., April 18. Of these, 67 individuals sustained low intakes of radioactive material.
(The highest exposure was about 5% of the 10 CFR 20 quarterly intake limits.)
i During the investigation, it was determined that the ventilation system installed to remove airborne radioactivity from the No. 22 steam generator was not in operation. During the work in the cold leg of the steam generator, air was to be drawn in through the cold leg, up through the steam generator tubes, I
out the cold leg manway, and finally exhausted into the containment atmosphere.
l The air was filtered by an iodine removal unit prior to its being exhausted i
into containment.
This removal unit, however, had not been turned on prior to the work on the No. 22 cold leg.
Review of general area airborne radioactivity sampling during the entire work activity indicated that samples were collected in the No. 22 steam generator tent and in the vicinity of the No. 22 reactor coolant pump.
It was noted, however, that no short duration grab samples were collected 'and analyzed.
In addition, no continuous air monitor (CAM) was operated in the vicinity.
The licensee did have one CAM operating in the containment.
However, the monitor was operating on the refueling floor and was later determined to be operating improperly.
As a result, the licensee had no timely indication of increasing airborne radioactivity.
. In addition, review of the licensee's procedure for the installation, repair, and removal of nozzle dams indicated limited guidance for minimizing radiation exposure to workers.
The procedure had been revised on April 14, 1983, to provide guidance for the removal of broken studs from the steam generator nozzle ring, but no guidance dealing with radiological control matters was included in the revision.
No process control signoffs dealing l
with radiological control matters, such as ensuring that the installed ventilation system was operable, were included in the procedure.
The licensce met with NRC staff in August 1983 to discuss the various problems and violations discussed above. Corrective actions by the licensee include revisions to several procedures involving assessment of airborne i
i radioactivity. The event and corrective actions remain under NRC review.
l (Refs. I through 3.)
1.2 Degraded High Pressure Safety Injection System On April 5,1983, the licensee for Maine Yankee
- reported to the NRC that the reactor had been operated between March 7 and April-5,1983 with unaware that one of the two high pressure safety injection (HPSI) plant operators subsystems was inoperable.
This subsystem was inoperable because its pump would not automat-ically start if called upon. An NRC inspector, during a subsequent review on April 6 and 7, determined that the other HPSI subsystem had also been inoperable with the plant at power for about seven hours on March 17, in that it would not have started if called upon during a loss of offsite power.
i At Maine Yankee, the HPSI pumps also function as the reactor coolant system charging pumps; one pump can provide sufficient charging flow with the plant at full power. The A or B pump can be functionally replaced by an installed spare pump, S.
The use of the spare pump allows preventive and corrective maintenance to be performed on either the A or B pump without the loss of i
redundancy.
In order to prevent operating two HPSI pumps from the same electrical distribution bus, the spare pump is prevented from automatically starting if the S pump supply breaker and either the A or B pump supply i
breaker are aligned to the same bus.
This interlock is accomplished by limit switches which actuate when either the A or B pump breaker elevator is in the i
raised position.
In December 1982, the A pump had been removed from service for overhaul of the pump motor, and was functionally replaced by the S pump.
Prior to the event. the S pump was last operated satisfactorily on March 3,1983 during routine surveillance testing.
Following the test, the S pump was placed in standby as the A train HPSI pump, and the B train HPSI pump was used as the reactor coolant system charging pump and the B train HPSI pump.
On March 7, at the request of maintenance personnel, plant operators tagged out the A pump (which was already out of service) to allow the pump motor to be cleaned.
In accordence with procedures, the breaker was removed and a grounding device f
was installed in the A pump breaker cubicle in the 4160 volt electrical distribution switchgear.
To install this device, the breaker elevator had to Maine Yankee is an 810 MWe (net) PWR located ten miles north of Bath, Maine, and is operated by Maine Yankee Atomic Power.
i 4
- __v,
r be lifted. As discussed previously, raising the breaker elevator actuated the interlock which prevented the S pump from automatically starting.
(The pump could have been manually started from a local control station.) At the time, neither the operators nor plant maintenance personnel realized that installation of the grounding device would cause the S pump to be inoperable.
The licensee was unaware that the system remained in the inoperable condition until April 5, when the S pump failed to start during routine monthly surveillance testing while the plant was operating at full power.
Within five hours, the licensee discovered the cause of the failure, removed the grounding device, and satisfactorily tested the S pump.
During a subsequent review of plant records on April 6 and 7,1983, the NRC inspector determined that from 8:30 a.m. until 3:15 p.m. on March 17, while the S pump was incapable of being automatically started, the licensee had removed diesel generator DG-1B from service for maintenance.
This rendered the HPSI B pump incapable of being operated by its emergency power source, had there been a loss of offsite power.
(Since no such loss of offsite power occurred, the B pump continued operating as a reactor coolant charging pump.)
During the approximately seven hours when the B diesel generator was out of service, automatic HPSI operation was unavailable in the event of loss of offsite power.
The window of vulnerability was small, particularly considering that the plant's technical specifications only require that if both HPSI trains are out of service, actions are begun within one hour so that the plant is in hot standby within six hours.
Although the S pump could have been manually started from a local control station if necessary, the plant was operated for a period of approximately one month while, unknown to the licensee, the S pump would not have started automatically in the event of a loss-of-coolant-accident. During this period of time, there was no redundancy for the HPSI function without operator action.
The following corrective actions have been undertaken by the licensee:
(1)
Specific procedural and administrative controls have been established for all safeguards pumps which incorporate the interlock design feature.
(2) A memorandum describing the new procedural controls was issued to the plant operators.
(3) An explicit warning sign was placed on all affected safeguards pump breaker cubicles.
(4) Operator and electrical maintenance training programs were revised.
(5) A design change is being developed to modify the S pump interlock such that it cannot be actuated other than as intended.
(6) The plant systems training manual is being revised to include tables of the electrical interlocks for all plant safeguards pumps.
The NRC determined that these events demonstrated the need for improvements in both procedures and training programs at Maine Yankee.
An enforcement
- )
conference between NRC Region I personnel and licensee representatives was held on April 21, 1983 to discuss the inspections performed and the violations identified. On May 20, the NRC issued a notice of violation and proposed imposition of civil penalty (for $40,000) to the licensee.
The licensee acknowledged the violation, described corrective actions, and paid the civil penalty on June 17, 1983.
(Refs. 4 through 7.)
A similar event occurred at Maine Yankee on August 19, 1983.
While operating at full power, the S pump failed to start following realignment to the A train HPSI pump.
Investigation revealed that a breaker cubicle interlock for the A pump, which had been lined up as the A train HPSI pump to conduct a performance test, remained engaged when the A breaker was not fully racked down during the process of realigning the S pump.
The A breaker had been racked down until it discharged, but its wheels remained approximately 1/8 inch off the floor.
Since the housing interlock prevents the A and S pumps from being powered by the same bus, after the S breaker was racked up it failed to close when actuated.
After the A breaker was fully lowered, the S breaker closed when activated and the S pump tested satisfactorily.
Plant operations personnel were instructed to ensure that breakers are completely lowered to properly activate housing limit switches.
Also, the licensee modified the interlock design so that the interlock is activated only by a fully racked up and charged breaker.
(Ref. 8.)
1.3 Service Water Bay Leak Causes Loss of Flow to Boron Injection, Residual Heat Removal, and Diesel Generator Systems On June 23, 1983, with Salem Unit 2* in cold shutdown, an equipment operator performing routine surveillance discovered a large leak in the No. 2 service water bay. The operator reported the leak to the control room.
Since the operator suspected that the header flexible coupling had failed, the operating pumps in the bay were deenergized and an attempt was made to isolate the leak by shutting the No. 21 nuclear header supply valve 22SW20. Due to the accumulation of approximately 6 feet of water in the bay and an apparently continuing rise in the level, all operating service water pumps were then stopped to protect the pump motors. However, crosstis valves 21SW23 and 22SW23, open due to the shutdown configuration, and backflow from the No. 22 header contributed to the flooding.
The loss of service water flow to the charging pumps, residual heat removal (RHR) pumps and heat exchangers, and diesel generator coolers rendered the associated systems and ac power circuits inoperable.
An unusual event was declared and appropriate notifications were made in accordance with emergency procedures.
The flooding was stopped before reaching the motors of the service water pumps.
The leak was isolated manually and service water flow was restored within an hour. Only limited damage to controls and equipment in the bay resulted.
No core alterations or positive reactivity additions were involved in the occurrence; containment integrity was being maintained at the time.
Salem Unit 2 is a 1106 MWe (net) PWR located 20 miles south of Wilmington, Delaware, and is operated by Public Service Electric and Gas.
' i Investigation revealed that the leakage was from the downstream flange of check valve 22SW5; the rubber gasket which was installed at the joint was found to have failed.
The connection had been reassembled following recent cleaning of the No. 21 service water header.
Some leakage had occurred following the reassembly and was eliminated by retightening the studs at the joint.
Some distortion of the gasket was observed around the stud holes in the vicinity of the failed portion of the gasket.. Installation and proper tightening i
of the connection is difficult due to physical obstructions hindering access to the rear and underneath the flange.
A controlled work package was used for the task, and included appropriate quality control verifications.
The failure was therofore assumed to involve an isolated problem in installation of the gasket.
The original sump pumps in the service water bay had previously failed.
A design change to install improved pumps was in progress, but had not yet been implemented.
Temporary pumps were being utilized in the interim.
Upon receipt of a sump high level alarm, an operator would be dispatched to start the temporary pump.
The alarm therefore did not provide warning of the size j
of a leak into the bay.
lne boron injection system, including the charging pumps, ensures that negative reactivity control is available during all modes of facility operation.
With the reactor coolant system (RCS) temperature below 200'F, only one injection system would be required on the basis of the stable reactivity condition of the reactor and the additional action requirements in the event the system becomes inoperable.
The operation of one RHR pump provides adequate flow to ensure mixing, prevent stratification and produce gradual reactivity changes during boron concentration reductions in the RCS. Single failure considerations require that the two RHR loops be operable in cold shutdown.
Finally, the operability of the emergency diesel generator, including the associated ac power circuits, during shutdown and refueling ensures that the facility can be maintained in this condition for extended time periods and that sufficient instrumentation and control capability is available l
for monitoring and maintaining the unit status.
i Service water was restored, and all safety-related equipment was restored to operation in a timely manner. The incident would not likely have occurred during operation at power, when redundant equipment would not be removed from operation for maintenance or isolation of the leak.
The service water headers are normally separated during power operation, minimizing the effect of a leak in one bay on the redundant header and leads. liowever, this event emphasizes the importance of maintaining operability and redundancy of safety-related systems during all modes of operation, since the combining of redundant service water systems during cold shutdown resulted in a single failure capable of rendering both service water trains inoperable.
The appropriate action was taken immediately to stop the' leak and insure compliance with the technical specifications. The flexible coupling at the joint was inspected and no problems were evident.
The new gasket was l
1 installed, the joint was reconnected, and the system was satisfactorily tested. No further problems have been noted with the joint.
Implementation l
of the design change for improved, permanent sump pumps will be expedited.
This will allow proper use of the alarm function for leak detection.
(Ref. 9.)
1.4 Failure of Service Water System Valves On February 9,1983, during performance of surveillance testing on Surry i
Unit 1,* four 30-inch normally closed butterfly valves on the service water system would not open upon demand from the control room.
These valves supply cooling to the heat exchangers for the containment recirculation spray system.
The Unit 1 reactor was shut down at the time with primary system temperature at 350'F and pressure at 450 psig.
Subsequent testing of the same valves on Unit 2, which was operating at 100% power, revealed that one valve opened fully, two valves failed in an intermediate position, and one valve failed to open.
The recirculation spray subsystem at the Surry Power Station is designed to mitigate the effects of a postulated loss-of-coolant accident (LOCA) during the recirculation phase of the recovery (i.e., after the water in the refueling water storage tank has been nearly depleted).
This is accomplished by returning the containment pressure to a subatmospheric value following a LOCA, providing for long term heat removal, and maintaining the containment at subatmospheric conditions.
The extended loss of the recirculation spray subsystem subsequent to a LOCA, therefore, could result in the loss of the ultimate heat sink.
Primary coolant water and emergency core cooling water, spilled from a postulated LOCA, and the water from the containment spray system, would collect in the containment sump.
While in the recirculation mode, the containment sump water would be diverted to two subsystems:
(1) the low head safety injection subsystem, and (2) the recirculation spray subsystem.
Water from the low head safety injection subsystem would be returned to the reactor to cool the core; however, there are no heat exchangers in that subsystem. Water from the recirculation spray subsystem would cool and depressurize containment by the heat removal function of the four recirculation spray coolers.
The recirculation spray coolers would not only cool the containment spray water but would also cool the water being returned to the reactor.
However, before these coolers can cool the containment and remove the decay heat, at least one of the four normally closed valves that admit service water to the coolers must be opened.
After the four Unit 1 valves failed to open, each was manually opened by using the valve's hand wheel.
Significant torque was required to lift the butterfly disks off their seats. The fact that each valve's failure to open resulted in the trip of its motor overload device indicates that significant torque would be required to open the valves. After the valves had been manually opened, two responded to the electrical signals from the control room and two failed to respond. As a consequence, the licensee was unable to determine the exact causes of the initial failures for those valves that subsequently responded to the electrical signals.
The cause of failures of those that did not respond Surry Units 1 and 2 are each 775 MWe (net) PWRs located 17 miles northwest of Newport News, Virginia, and are operated by Virginia Electric and Power.
- _. to the electrical signals was believed to be corrosion (one had a corrodeo motor, the other a corroded torque switch). Although the exact cause (or causes) for these failures has not been determined, they have been attributed to several factors, or a combination thereof, including:
. Marine growth - barnacles, shells, silt and seagrowth were found on the valve seats and disks;
. Corrosion - two valves failed to open because of a corroded motor and another because of a corroded torque switch.
The corrosion appears to be due to either the flooding of the valve pit area that occurred several years ago or to the high moisture content in the valve pit area;
. Infrequent testing - these valves had been only stroke-tested during refueling outages (i.e., approximately every 18 months);
. Low torque switch settings - two Unit 2 valve operatcrs " torqued out" when tested; and
. Marginally sized motors and/or improperly geared valve operators.
Corrective actions taken by the licensee include:
. Cleaning the piping and valve internals;
. Applying marine inhibitor paint on the valve disks;
. Testing the valves quarterly rather than during refueling outages.
(Note:
The Unit 1 valves operated properly during subsequent testing, but one Unit 2 valve failed to open during subsequent testing because of a low torque switch setting);
. New motor operators having increased torque output have been engineered, and one such operator is being tested at the plant before installing eight such operators.
In the interim, the torque switch settings of the installed operators have been increased.
. Procurement procedures for the purchase of new valves and operators have been initiated; and
. Administrativa controls have been revised so that responsibility for the l
proper operation of these valves has been specifically assigned to a knowledge-able person on each shift. This person is responsible for ensuring the proper operation of these valves during accident conditions by going to the valve pit area and manually opening the valves, if so required.
In addition, prior to this event but subsequent to the above mentioned flooding of the valve pit area, metal dikes had been constructed around the valve pits and water level detectors had been installed to prevent or detect flooding of these valves.
(Refs. 10 and 11.)
1.5 Reactor Yessel Internals Bolting Failures Most operating Babcock & Wilcox (B&W)-designed pressurized water reactor (PWR) plants have experienced various degrees of degradation of nickel / chrome steel bolts used in fastening the reactor vessel internals.
The currently operating B&W-designed plants are listed below. Of these plants, only Davis Besse Unit 1 showed no evidence of bolt degradation during inspections in 1982 and August 1983.
Plant Licensee Location Arkansas Nuclear One Arkansas Pot =r and Light 6 miles WNW of Russellville, Unit 1 Company Arkansas 836 MWe (net)
Crystal River Unit 3 Florida Power CorporaHon 7 miles NW of Crystal River, 782 MWe (net)
Florida Davis Besse Unit 1 Toledo Edison Company 20 miles E of Toledo, Ohio 874 MWe (net)
Oconee Units 1, 2, 3 Duke Power Company 30 miles W of Greenville, 860 MWe (net) each South Carolina Rancho Seco Sacramento Municipal 25 miles SE of Sacramento, 873 MWe (net)
Utility District California The bolts affected are made of A286 material (a form of nickel / chrome steel) and are located in the upper and lower core barrel, upper and lower thermal shield, and the flow distributor to lower grid.
The bolts have welded locking clips to capture the bolt and prevent bolt rotation.
In addition, Davis Besse Unit I and Crystal River Unit 3 have surveillance specimen holder tubes bolted to the thermal shield.
Of these joints, only the upper and lower core barrel bolting have core support significance.
Should all the bolts in either of these joints fail, the reactor core and internals would drop onto guide lugs welded to the inside wall of the reactor vessel.
These lugs are designed to limit the core drop to about 0.5 inch.
B&W analyses have shown that reactor shutdown and subsequent coolability are assured even if all of the bolts in either core barrel joint fail.
The reactor coolant pressure boundary is also maintained.
However, such a situation would represent a significant reduction in the intended design safety margin; there-fore, the integrity of the core barrel bolted joints must be assured.
Cracking in the A286 bolting material was first noted at Oconee Unit 1 during 1981 in the bolts in the lower thermal shield. As a result, a program for reactor vessel bolt inspection and repair was initiated. Ultrasonic (UT) inspection results, showing the number of cracked bolts out of the number inspected, are shown in Table 1.
As a result of the discovery of the core barrel bolt anomalies at Rancho Seco and Crystal River Unit 3, the B&W Owners Group formed a special task force on internals bolts. The task force was to provide an initial evaluation of the degraded bolt situation.
The task force's findings showed that operating facilities provided no undue risk to public health and safety.
In addition, the task force was to continue to investigate and evaluate new information for resolution of problems found. The results of the initial evaluation were presented to the NRC by Task Force report BAW 1784 dated May 6,1983, which was formally submitted by a B&W letter dated May 20,1983 (Refs.12 and 13).
Table 1 UT Inspection Results (# Cracked /f Inspected) j Crystal River Bolted Joint Oconee Unit 1 Oconee Unit 2 Oconee Unit 3 Arkansas Unit 1 Rancho Seco Unit 3 j
Date of Inspection 9/81 1/8?
6/52 5/83 3/83 4/83 Upper Core Barrel 0/21 0/30 0/30 7/120 19/120 51/120 (120 - 1/3/4" Dia.)
0/108 4/108 Lower Core Barrel 0/16 0/24 0/24 (108 3/4" Dia.)
0/93 0/96 Flow Dist/ Lower Grid 0/22 0/25 0/25 i
(96 - 1" Dia.)
0/60 0/60 Upper Thermal Shield 0/25 0/20 0/20 (60 1/2" Dia.)
Lower Thermal Shield 11/13 28/93 53/96 51/96 77/96 74/96 I
(96 - 1" Dia.)
94 heads Several heads Several heads 48 heads 75 heads 71 heads twisted off twisted off twisted off twisted off twisted off twisted off 25/72 Surveillance Holder i
Tube *
(72 - 3/4" Dia.)
Davis Besse Unit 1 performed visual inspections during their 1982 refueling outage beginning in March 1982, and found.no abnormal indications.
Also, no abnormal indications were found during inspections in August 1983.
Davis Besse Unit 1 and Crystal River Unit 3 only.
l l
Laboratory examinations of some bolts from Oconee Units 1 and 2 and Rancho Seco appear to indicate that bolt failures in the reactor vessel internals seen to date are due to intergranular stress assisted cracking located in the bolt head to shank transition region. As a permanent fix to the problem, Oconee has reported that they have replaced all 96 bolts and locking clips on the lower thermal shield with a stud and nut assembly design.
Thermal shield bolt repairs have been completed on all of the affected plants.
Oconee has reported that the Unit 1 upper core barrel bolts, ten lower core bolts, and 11 thermal shield studs were 100% ultrasonically tested during a June-July 1983 outage, and showed no evidence of degradation.
As previously mentioned, Davis Besse Unit I showed no indications of bolt degradation when the bolts were inspected in 1982 and August 1983. All upper core barrel bolts were replaced in Crystal River Unit 3 and Rancho Seco.
In addition, two upper and four lower core barrel bolts were removed in Rancho Seco for testing / archival purposes.
The B&W Owner's Group Task Force performed analyses to determine the minimum number of core barrel bolts needed to maintain joint integrity, both for normal operation loads and for faulted condition loads (including loads generated by a large break loss-of-coolant accident and core bounce, a safe shutdown earthquake, and the dead weight of the core and internals).
Depending upon the particular plant, only from 7% to 23% for normal operations, and 36%
to 42% under faulted conditions, of the 120 upper core barrel bolts are required to maintain integrity.
For the 108 lower core barrel bolts, the corresponding numbers are 4% to 22%, and 15% to 20%, respectively.
The affected licensees submitted justifications for continued plant operation, until their next scheduled refueling / maintenance shutdowns, based on the inspections, analyses, and.any repairs made to date.
The justifications were further based on (a) the number of core barrel belts utilized in the designs result in large structural design margins, (b) severe bolt f ailure is generally detectable by available instrumentation, (c) the consequences of core barrel joint failure do not constitute a significant reduction in public health or safety status, and (d) the bolt degradation is caused by slowly-developing mechanisms, with failed bolts being generally randomly distributed.
The Task Force is continuing their investigations toward understanding the cause and solution of the problem.
The NRC has agreed that the licensees could continue plant operation where justifications for continued operation have been submitted, or where inspec-tion results showed no degradation of bolting.
I 1.6 Events Involving Two or More Simultaneous Dropped Rod Control Cluster Assemblies l
The following information concerns seven events which involved the dropping l
of two or more rod control cluster assemblies (RCCAs).
The most recent i
event occurred at Turkey Point Unit 3 on March 9,1983. As a result of this event, a data base system search was conducted, covering the approximate period l
from January 1976 to March 1983.
This search identified six additional events involving the dropping of multiple rod control cluster assemblies.
The seven events are described below.
' i l
Turkey Point Unit 3 l
On March 9,1983, wnile Turkey Point Unit 3* was operating at 100% power, rod control cluster assemblies 08 and M8 of control bank D simultaneously dropped.
j The applicable technical specification requires that sustained power operation of 1
i the unit shall not be permitted with more than one inoperable control rod.
The nuclear instrument system and the rod position indication dropped rod circuitry caused an automatic turbine runback to 70% power.
According to plant procedures, power reduction to hot shutdown was initiated.
Within approximately 15 minutes after tha shutdown began, the two dropped rods were retrieved and were verified to be back in the correct position by a flux map.
Immediate inspection of the control rod power cabinet revealed that water was dripping on the cabinet and seeping inside.
It appears that the water inside the cabinet resulted in a momentary short in the lead to the D8 and M8 control rod stationary coils.
The cabinet was dried and the water leakage was stopped at the source. The unit was subsequently returned to full power.
Westinghouse (the nuclear steam supply system vendor) was consulted, and it was determined that no safety limits were encroached upon due to the short duration of the incident and the subsequent operability of the two rod control cluster assemblies.
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Because the simultaneous dropping of two control rods was considered to be a rare event, Turkey Point Unit 3 did not have explicit plant procedures addressing it.
The plant procedures at this station mainly addressed operator action in the case of one dropped rod.
However, the March 9 occurrence has made plant personnel aware of the need to address the multiple-dropped-rod incident in the operating procedures.
For this activity, Westinghouse will be consulted for guidance in writing new immediate actions for the multiple-dropped-rod incident.
St. Lucie Unit 1 On September 4,1980, while changing the ISV power supply for control element assembly (CEA) 42 at St. Lucie Unit 1,** CEA 44 dropped, possibly due to a voltage spike. Due to a failure of the public address system, maintenance personnel were not notified that CEA 44 dropped, and they continued removing the power supply for CEA 42 which led to the dual rod drop.
Action in accordance with technical specifications was taken, and the reactor was manually tripped. No abnormal subsequent actions or occurrences were identified.
Turkey Point Unit 3 is a 646 MWe (net) PWR located 25 miles south of Miami, Florida, and is operated by Florida Power and Light.
St. Lucie Unit 1 is an 817 MWe (net) PWR located 12 miles southeast of Ft. Pierce, Florida, and is operated by Florida Power and Light.
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_ Calvert Cliff s Unit 2 On April 17, 1982, during normal operation at Calvert Cliffs Unit 2,* CEA 21 dropped into the core after electricians inadvertently disconnected power from it.
Approximately three minutes later, CEA 20 dropped into the core for the same reason, at which point the reactor was manually tripped.
The cause of this event was attributed to personnel error.
To preclude recur-rence, the lead person on the job was interviewed by supervision and instructed in the proper means of determining job scope.
A revision to an administrative instruction also was made.
Arkansas Nuclear One, Unit 2 On December 6, 1978 at Arkansas Nuclear One Unit 2,** during low power physics testing in the startup mode, CEAs 26, 1, 48, 46, 57, and 52 dropped and CEAs 22, 27, and 29 slipped while moving for positioning.
The individual CEA misalignments caused no consequences since the reactor power was less than 17..
The dropping or slipping of CEAs 26,1, 48, 46, 57, 22 and 52 was attributed to timer card failures, whereas the slipping of CEA 27 and 29 was attributed to sequencer card failure.
Subsequent to these occurrences, repairs were made and operations continued.
Ginna The remaining three events occurred at Ginna.***
No abnormal system behavior or unexpected actions were identified during or following any of these events.
On August 4,1976, during steady load, turbine runback occurred when D bank group 2 rod control cluster assemblies G-3 and G-11 partially dropped into the core. Although the cause of the two dropped rod control cluster assemblies was not explicitly identified, it was believed to be associated with the 2BD power cabinet, which provides power to these two rod control assemblies.
A similar event occurred on July 4,1976. During steady load, turbine runback occurred when B bank group 2 rod control cluster assemblies G-5 and G-9 dropped.
The exact cause for the dropping was not identified; however, on two previous occasions dropping of these two rod assemblies was associated with water leaks over the attendant power cabinet.
An alarm card indicated stationary coil A regulation failure.
Following this occurrence, and based on a recommendation by Westinghouse (the nuclear steam system vendor), fuses and current sensing resistors were replaced in the stationary coil circuitry.
Calvert Cliffs Unit 2 is an 825 MWe (net) PWR located 40 miles south of Annapolis, Maryland, and is operated by Baltimore Gas and Electric.
Arkansas Nuclear One, Unit 2 is an 858 MWe (net) PWR located 6 miles west northwest of Russellville, Arkansas, and is operated by Arkansas Power and Light.
Ginna is a 470 MWe (net) PWR located 15 miles northeast of Rochester, New York, and is operated by Rochester Gas and Ele
_.. _ ~ __
_ _ _ i An event on June 16, 1976 involved the dropping of B bank group 2 control rod cluster assemblies G-5 and G-9.
For this event, the alarm card detector lamp indicated four printed circuit cards as the possible cause, however, j
no explicit cause was identified.
)
The safety implications of the seven multiple rod drop events described above are minor, since the rate of these events is infrequent (about one i
per year, based on the data base set of events) and the consequences are j
as expected.
Fuel design limits were not exceeded in any of the events.
The causes of the events appear to be the direct result of either maintenance personnel error and/or disturbances in the attendant rod control cluster assembly power supply circuitry. Also, it should be noted that the operating procedures for at least one of the plants identified above did not explicitly address the multiple dropped rod incident.
(Refs.14 and 15.)
1.7 Damage to Thermal Shield On March 30, 1983, the NRC was notified that loose metal parts had been i
found inside the reactor vessel of the St. Lucie Unit 1* facility.
These pieces of metal were later determined to be parts from the thermal shield support, alignment, and positioning systems.
E During a refueling outage, while loading fuel, loose metal parts as well as j
metallic debris were found inside the reactor vessel.
The larger pieces found appeared to be locking bars for thermal shield positioning pins and pieces of the thermal shield. The reactor was defueled, and the core barrel and thermal shield assembly were removed for inspection.
Damaged or missing l
thermal shield support lugs, upper positioning pins, and lower alignment pins were identified. Damage was also identified on the core barrel in the vicinity i
of the core barrel to thermal shield support lug welds.
The reactor vendor, Combustion Engineering (CE), conducted a review of the worst case that could have resulted had the thermal shield support mechanism been damaged to the extent that the thermal shield dropped.
It was determined that, in this case, the fall of the thermal shield would be arrested, with little impact on flow through the core.
The licensee has determined that no fuel damage occurred as a result of the thermal shield damage and, additionally, CE concluded that in the worst case, were fuel damage to occur, other instrumentation would detect such damage prior to the occurrence of a significant failure.
It is believed that the damage to the thermal shield was not a single event l
but rather occurred over a period of time and was related to mechanical i
stress caused by flow inouced vibrations. The licensee for St. Lucie Unit 1 l
plans to remove the thermal shield from the core barrel and to continue operation
(
without the thermal shield.
The licensee has analyzed the consequences of this action and found that the increased neutron flux seen by the reactor.
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vessel for operation without the thermal shield would present no problems with respect to the effect on pressurized thermal shock or embrittlement of the reactor vessel.
St. Lucie Unit 1 is a 777 MWe (net) PWR located 12 miles southeast of Ft. Pierce, Florida, and is operated by Florida Power and Light.
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- Other CE licensees with thermal shields
- have analyzed the implications of this event and submitted justifications for continued operation to the NRC.
The NRC is studying the results of analyses performed by the licensee and CE that discuss the safety significance of the damage to the thermal shield, and the implications of continued operation without the shield. The NRC has also conducted onsite inspections and has held meetings with the licensee, CE, and the other affected licensees.
The NRC is continuing to review this event as well as the progress of the licensee's corrective actions at St. Lucie Unit 1.
(Refs. 16 and 17.)
Only four CE plants have thermal shields; two of these are removing their shields.
I 1.8 References (1.1) 1.
NRC, Preliminary Notification PN0-I-83-32, April 18,1983.
2.
NRC, Inspection Report 50-311/83-14, June 8,1983; and Notice of Violation for Docket 50-311, Licensu DPR-75, June 28,1983.
3.
Letter from T. Martin, NRC/wgion I, to R. Uderitz, Public Service Gas and Electric Company, CAL-83-12, August 18, 1983.
(1.2) 4.
Maine Yankee Atomic Power Company, Docket 50-309, Licensee Event Report 83-09, April 25,1983.
5.
NRC/ Region I, Inspection ReccH. 50-309/83-05, April 19, 1983.
6.
Memorandum from J. Allan, NRC/ Region I, to J. Randazza, Maine Yankee Atomic Power, May 20, 1983.
7.
Letter from J. Randazza, Maine Yankee Atomic Power, to R. DeYoung, NRC, June 17, 1983.
8.
Maine Yankee Atomic Power Company, Docket 50-309, Licensee Event Report 83-30, September 21, 1983.
(1.3) 9.
Public Service Electric and Gas Company, Docket No. 50-311, Licensee Event Report 83-32, July 13,1983.
(1.4)
- 10. Virginia Electric and Power Company, Docket No. 50-280, Licensee Event Report 83-03, February 23, 1983.
11.
NRC, Information Notice No. 83-46, July 11, 1983.
(1.5)
- 12. Babcock & Wilcox (B&W), Owners Group Bolting Task Force Report, BAW 1784, May 1983.
- 13. Letter from J. H. Taylor, B&W, to S. Miner, NRC, May 20, 1983.
(1.6) 14.
NRC, AE0D Technical Review Report T324, March 9,1983.
15.
NRC M2morandum from L. Rubenstein, NRR/DSI, to F. Miraglia, NRR/DL, March 2,1983.
(1.7) 16.
NRC, Preliminary Notification PN0-II-83-18A, April 6,1983.
17.
Florida Power and Light Company, Docket No. 50-335, Licensee Event Report 83-22, April 19,1983.
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.
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l 2.0 ABSTRACTS OF OTHER NRC OPERATING EXPERIENCE DOCUMENTS 2.1 Abnormal Occurrence Reports (NUREG-0090) Issued in May - June 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 deter. nines is significant from the standpoint of public health or safety. Under the provisions 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 t
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 Report S/83 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES: OCTOBER - DECEMBER 1982, NUREG-0090, VOL. 5, NO. 4 During the report period (on October 28,1982) one abnormal occur-rence was reported by an NRC licensee, Alabama Power Co.
This event involved an inoperable containment spray system at Farley Nuclear Plant Unit 2.
i In this fourth calendar quarter of 1982, neither fuel cycle facilities (other than nuclear power plants), nor other NRC licens?es (e.g., industrial radiographers, medical instituticns, etc.), nor Agreement States reported any abnormal occurrences to the NRC.
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. _ _ _ _ _ _ _ 2.2 Bulletins and Information Notices Issued in May - June 1983 The Office of Inspection and Enforcement periodically issues bulletins and information notices to licensees and holders of construction permits.
During the period, one bulletin and 18 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 enforcement action, such as an order for suspension or revocation of a license. When apprepriate, 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 intended as substitutes for revised license conditions 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 acknowledgment or response, but recipients are advised to consider the applica-bility of the information to their facility.
Date Bulletin Issued Subject 83-05 5/13/83 ASME NUCLEAR CODE PUMPS AND SPARE PARTS MANUFACTURED BY THE HAYWARD TYLER PUMP COMPANY All nuclear power reactor licensees and construction permit holders were apprised that the Hayward Tyler.
Pump Company (HTPC) had shipped certain defective safety-related pumps to various domestic and foreign nuclear plants. ' Licensees who use or plan to use ASME Csde pumps and spare parts manufactured by HTPC during the period 1977 to 1981 should conduct pump performance / endurance tests to ensure pump reliability and should provide NRC a summary of the inservice test requirements or plans to develop such requirements for affected pumps. Those facilities using the HTPC pumps in question, the pump service, and number.of pumps were listed in an attachment.
The performance tests recommended by HTPC were provided, as were additional recommendations for installation of replacement parts.
, Information Date Notice Issued Subject 83-26 5/3/83 FAILURE OF SAFETY / RELIEF VALVE DISCHARGE LINE VACUUM BREAKERS All nuclear power reactor facilities holding an operating license or construction permit were provided early notification about the failure of installed vacuum breakers on safety / relief discharge lines at Browns Ferry Unit 1 and Peach Bottom Unit 2 (boiling water reactors).
At Browns Ferry, the licensee found a damaged hinge pin in the leaking vacuum breaker (10" GPE check valve), replaced it, the solenoid, and the pilot cartridge, as well as four other safety relief valve (SRV) vacuum breakers that were damaged.
An SRV that had leaked on February 5,1983 and had been replaced, leaked again three days later and again was found to be damaged.
At Peach Bottom, two vacuum breakers (8" Anderson Greenwood check valves) were replaced when they were damaged during an inspection. Both showed some binding on the hinge pin.
The breakers that failed had been installed for SRV second-pop transient protection.
Some recently licensed plants are using the low-low setpoint relief function in lieu of vacuum breakers for such protection.
83-27 5/4/83 OPERATIONAL RESPONSE TO EVENTS CONCERNING DELIBERATE ACTS DIRECTED AGAINST PLANT EQUIPMENT l
All nuclear reactor facilities holding an operating license or construction permit were notified about deliberate acts of tampering with plant equipment involving improper valve positioning and instrumenta-tion irregularities.
In the cases described, licensees were not totally prepared to assess the situation and to respond so as to ensure that systems important to safety remained operable and to make decisions about continued operation. The facilities were advised to provide guidelines and station procedures to cope with such threats to safety in accordance with 10 CFR 73.55(h)(1) and Appendix C of Part 73.
83-28 5/4/83 CRITERIA FOR PROTECTIVE ACTION RECOMMENDATIONS FOR GENERAL EMERGENCIES All nuclear power reactor facilities holding an operating license or construction permit were reminded that emergency plans and procedures are required
Information Date I
Notice Issued Subject to contain site-specific protective action recommenda-tion guidelines as provided in Appendix 1 of NUREG-0654/ FEMA-REP-1, Rev.1, " Criteria for Preparation and Evaluation of Radiological Emergency Preparedness in Support of Nuclear Power Plants."
83-29 5/6/83 FUEL BINDING CAUSED BY FUEL RACK DEFORMATION 4
All nuclear power reactor facilities holding an operating license or construction permit and spent fuel-storage facilities were notified about problems L
at Yankee Rowe (1963), Haddam Neck (1978), Kewaunee (1980), and Maine Yankee (Oct.1982), involving 1
cell bulging in spent fuel racks or hydrogen liberated in the spent fuel pool.
The deformation at Maine Yankee was so severe that it caused the fuel assembly to bind upon being inserted into the cell. All of the 21 deformed cells at Maine Yankee were " Phase I design," fabricated before 1975.
The.
licensee plans to drill and vent all affected cells as an interim fix; and then to replace them.
1 83-30 5/11/83 MISAPPLICATION OF GENERIC EMERGENCY OPERATING i
PROCEDURES (E0P) GUIDELINES All nuclear power reactor facilities holding an i
operating license or construction pennit were notified about the potential. for_' misapplying emergency operating procedures by not taking into.
account the operating modes for which those procedures were designed and conditions for which they apply.
Two inadvertent safety injection (sis) occurred at the Sumer station in March 1983.
The first occurred while the reactor was being cooled down and the -
operators, constrained by the licensee's Emergency Operating Procedure (EOP), did not terminate safety injection flow until the RCS pressure reached the 2000 psig constraint'of the generic technical guidelines, which were not intended for application during reactor cooldown, although plant procedures did not differentiate for this ' condition. - The second SI occurred on the next day.
This time, s
identifying the 'SI as being spurious, the reactor
-operator did not wait for the E0P termination
- criterion _of 2000 psig, but acted promptly to -
terminate SI.
j 83-31 5/19/83
-ERROR IN THE ADLPIPE. COMPUTER' PROGRAM L
All nuclear power' reactor facilities holding an operating license or ' construction permit, nuclear
. _ _ _ _. Information Date Notice Issued Subject steam system suppliers, and architect-engineers were notified about an error in the ADLPIPE computer l
program used for piping analysis.
The error affects only calculations using the Class 2 component option for the 1972 version of the ASME Code and does not affect calculations performed using the options for the 1974 and 1977 editions. The error is detailed and potential users of the program were informed how to correct the error.
83-32 5/26/83 RUPTURE OF AMERICIUM-241 SOURCE (S) CONTAINED IN A WELL LOGGING DEVICE All NRC licensees holding a specific license to possess and use sealed sources contaie,ing byproduct or special nuclear material in well iogging tools were alerted to how the integrity of sealed sources can be jeopardized by procedures to recover well logging tools stuck in the drill hole.
Licensees were advised to review these procedures to ensure that drilling or hole enlargement is not permitted during such operations until a clean break of the wireline is made at the point of attachment to the stuck device ensuring no cable remains attached to the radioactive source that could possibly draw it up in the shaft.
NRC also suggests that licensees ensure that the well head and/or mud discharge be continuously monitored with a suitable survey instru-ment or logging tool (minus the source) to immediately alert the operator to a possibly ruptured source so that contamination can be properly controlled.
83-33 5/26/83 NCNREPRESENTATIVE SAMPLING 0F CONTAMINATED OIL All nuclear power reactor facilities holding an operating license or a construction permit were reminded that it is difficult to obtain a representa-tive sampling of radioactive contamination from oil-water mixtures.
Licensees were advised that the contamination may be present in only the oil or only the water fraction, thereby potentially causing misleading results even in well mixed samples.
Emphasis should be placed on (1) minimizing, if not eliminating, the amount of water in waste-oil mixtures; and (2) testing waste-oil sampling techniques to confirm that representa-tive samples are being obtained.
Information Date Notice Issued Subject 83-34 5/26/83 EVENT NOTIFICATION INFORMATION WORKSHEET All nuclear power facilities holding an operating license or construction permit were provided with an event notification worksheet in order to facilitate rapid communication with the NRC Operations Center in the early stages of an incident at an NRC licensed facility.
It was suggested that all licensees copy the worksheet and make it available for reporting events.
83-35 5/31/83 FUEL MOVEMENT WITH CONTROL RODS WITHDRAWN AT BWRS All boiling water reactor facilities holding an operating license or construction permit were notified about potentially significant events involving fuel loading in control cell locacions where control rods are not fully inserted. Technical specifications at Brunswick Unit 1 and Duane Arnold allow fuel movement with multiple control rods withdrawn when specific conditions are met. Fuel movements at Brunswick Unit 1 had been made in accordance with the as-written procedure, but, in violation of technical specifications (TSs),
apparently were not reviewed to ensure that the control rods were fully inserted in those cells to which fuel was moved.
At Duane Arnold, also in violation of TSs, a temporary change in a fuel movement procedure inadvertently eliminated a step to insert the control rod in the next cell to be loaded.
NRC plans to reevaluate the TSs governing fuel movements, when multiple control rods have been withdrawn.
83-36 6/9/83 IMPACT OF SECURITY PRACTICES ON SAFE OPERATIONS All nuclear power reactor facilities holding an operating license or construction permit were informed about the potential for an adverse safety impact that exists at licensed facilities during abnormal or emergency conditions if plant operators are unable to pass through locked doors quickly because of computer failure, operator error, or procedural requirements.
Suggestions are offered to ensure that security practices do not inhibit safe operations.
I 83-37 6/13/83 TRANSFORMER FAILURE RESULTING FROM DEGRADED INTERNAL l
CONNECTION CABLFS 1
All nuclear power reactor facilities holding an operating license or construction permit were notified about a loss of all offsite power at Brunswick Unit I which resulted in the declaration of an Unusual Event.
The loss of power resulted
_ Information Date Notice Issued Subject from the inadvertent tripping of one of the two feeds from the station auxiliary transformers during performance testing.
The transformer feeding emergency bus E-6 in Unit 2's reactor building overheated, causing the feeder breaker to trip on overcurrent.
This trip was not created by the loss of power, but failure was attributed to improper assembly of transformer winding tap cables and long-time, undiagnosed, heat-induced degradation.
The transformer was manufactured by the ITE in 1972 and is a dry tape, 4160-volt primary (Delta) to 480/277-volt secondary (Wye).
83-38 6/13/83 DEFECTIVE HEAT SINK ADHESIVE AND SEISMICALLY INDUCED CHATTER IN RELAYS WITHIN PRINTED CIRCUIT CARDS All nuclear power reactor facilities holding an operating license or a construction permit were notified about failure of the adhesive that bonds heat sinks on loop power supply (NLP) printed circuit cards. Suspect heat sinks can be identified by inspecting the NLP card for hex nuts visible on the top side of the assembly. A new design heat sink has screw heads visible from the top side of the assembly. A secon,d problem has been contact bounce in the mercury relay utilized on temperature channel test (NTC) printed circuit cards.
(Westing-house is developing and testing a replacement relay.
for these NTC cards.) Both cards are produced by the Industry Electronics Division (IED) of Westinghouse.
The suspect cards are limited to those shipped from IED between August 1,1980 and September 1,1982, which are designated SNLP Sub-level 18 and above and 6NLP Sub-level 18 and above.
Domestic plants that use suspect cards are listed.
83-39 6/17/83 FAILURE OF SAFETY / RELIEF-VALVES TO OPEN AT BWR -
INTERIM REPORT All nuclear power reactor facilities holding an operating license or construction permit were pro-vided with an update of Information Notice No. 82-14 (see Power Reactor Events, Vol. 4, No. 4, p. 23). An owners group is funding the GE test program on Target Rock two-stage safety / relief valves (SRVs) to find out why SRVs specified to open within +1% of their setpoint failed to actuate. Of'seven Ihat failed to ectuate at 103% of set pressure, five showed signs of labyrinth seal friction and two had indications that a stuck pilot disk / seat condition existed.
Metallurgical examination of. one of the latter
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. Information Date Notice Issued Subject two valves and two other valves indicated corrosion products on the stellite alloy disk. As a result of this test and other tests, it was determined that leakage is unrelated to the problem of setpoint drift.
At Fitzpatrick Nuclear Plant, the reactor scrammed from 89% power, caused by a main steam isolation valve closure and recirculation pump trip.
The K relief valve (setpoint 1090 psig) did not lift.
At 1120 psig, the J relief valve (setpoint 1140 psig) lifted.
Since the event resulted in a relatively rapid transient, the K SRV failure tends to contradict the hypothesis that SRV sticking occurs only during slow pressure increases.
83-40 6/22/83 NEED TO ENVIRONMENTALLY QUALIFY EP0XY GROUTS AND SEALERS All nuclear reactor facilities holding an operating license or construction permit were alerted to potential degradation of epoxy formulations by heat and radiation. At Watts Bar, two epoxy grouts used to install anchor bolts inside the containment building showed significant loss of strength at temperatures above 120*F.
Thermal agiing, radiation environment, and the relatively low creep strength of epoxies must be considered when specifying epoxy formulations. Confidence in epoxies can only be achieved by product verification.
83-41 6/22/83 ACTUATION OF FIRE SUPRESSION SYSTEM CAUSING IN0PER-ABILITY OF SAFETY-RELATED EQUIPMENT All holders of operating licenses or construction permits were alerted to some experiences in which automatic actuation of fire suppression systems damaged or jeopardized the operability of systems important to safety.
It appears that errors have been made in design, in installation, and in plant operating and maintenance procedures. The events reported indicate that a walk-down of plant equipment would have identified instances where minor modifi-cations would have reduced water damage without significantly reducing the systems' effectiveness.
Overall design of fire suppression systems must provide an effective fire protection system but not adversely affect other aspects of plant safety.
83-42 6/23/83 REACTOR MODE SWITCH MALFUNCTIONS All boiling water reactors holding an operating license or construction permit were notified about reactor mode switch malfunctions while changing
I Information Date Notice Issued Subject operating modes. Following completion of a surveillance test at Susquehanna Unit -1 on March 22, 1983, the mode switch (:nodel name RUDEL-GOULD) was returned to the shutdown position from startup, but produced only a half scram signal.
The mode switch and a replacement switch exhibited numerous contact positioning i
errors that appear to be the result of significant j
irregularities in the mode switch cam shaft parts 4
and large design clearances resulting in imprecise operation of the cam followers.
Four switches j
passed a bench test and one of these was used as a replacement.
Less than two months later, a recently installed modified version of the type of mode switch that malfunctioned in the March 22 event malfunctioned at Susquehanna Unit I while changing from refuel to startup.
On December 17, 1982, a mode switch (SB-1) at Dresden Unit 2 failed when it was moved from the run position to the startup position.
Similar events had occurred previously at Dresden Unit 2 I
and at Quad Cities Unit 1 involving mode switch SB-1.
All had resulted in group 1 isolations from a low main steam line pressure signal after the switch was placed in a position that normally bypasses this signal.
It appears that because of years of service, the SB-1 switch has sticking contacts.
BWR owners should be alert to unexplained isolation or scram signals that may have coincided with mode switch movement while changing operating status.
It may be prudent for licensees to require the use of the scram pushbuttons before moving the mode switch to the shutdown position.
83-43 6/24/83 IMPROPER SETTINGS OF INTERMEDIATE RANGE (IR) HIGH FLUX TRIP SE1 POINTS All nuclear power facilities holding an operating license of a construction permit were informed of I
the need to re-compute trip values of intermediate i
range (IR) high flux trip channels if the ccre has been reloaded with a fuel configuration that has a low neutron leakage. At Morth Anna Unit 2, trip values 4 x 10-4 amperes had been equivalent to 25% power for the previous core pattern, but were too high for the new low neutrcn leakage fuel load pattern which requires 2.5 x 10-4 mperes for the'IR a
l high flux bistables to trip at 30% power or less.
A review of North Anna Unit I revealed a similar-
.. Information Date Notice Issued Subject situation. Zion Unit 1, Maine Yankee, and Surry I
Units 1 and 2 had conditions similar to those at North Anna subsequent to reconfiguring the cores to low leakage patterns.
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___- 2.3 Engineering Evaluations and Case Studiec Issued in May - June 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 imediate, 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 potentia 1) illustrated by the event and recommending actions to improve safety and 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 AE0D 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.
Engineering Date Evaluation Issued Subject E312 5/18/83 OPERABILITY OF TARGET ROCK SRVS IN THE SAFETY MODE WITH PILOT VALVE LEAKAGE During a return to power in July 1982 FitzPatrick operating personnel observed that one Target Rock two-stage safety-relief valve (SRV) tailpipe temperature was considerably higher than normal.
When disassembled the "F" SRV pilot valve was found to be eroded and leaking.
Similar. pilot valve leakage had been observed in two two-stage SRVs at Pilgrim 1 in 1981.
Two concerns related to safety mode operability were considered as a result of the pilot valve
(
leakage: upward setpoint drift at low leak rates and inoperability at high leak rates. Subsequent investigation by Target Rock indicated setpoint drift was not a problem for leak rates which did not challenge SRV operability in the safety mode.
No case studies were issued during May - June 1983, i
______ Engineering Date Evaluation Issued Subject f
Target Rock informed FitzPatrick that SRV safety mode operability would be challenged by leakage rates over 200 pounds per hour.
The principal findings of this evaluation are that current regulatory requirements do not adequately address SRV inoperability because of pilot valve leakage and that the limit on pilot valve leakage is not identified in existing regulatory guidance (i.e., technical specifica-tions, generic letters, IE Bulletins, etc.).
E313 6/15/83 P0TENTIAL CONTAMINATION OF THE SPENT FUEL P0OL AND PRIMARY REACTOR SYSTEM A previous engineering evaluation report, E242 (see Power Reactor Events, Vol. 4, No. 6, p. 24),
identified potential safety issues concerning stress corrosion cracking of the top nozzle of fuel assemblies in the spent fuel pool at Prairie Island. This evaluation provides an assessment of the information obtained as part of the program office investigation performed in response to E242.
Based on a review of the information, it appears that the failure mechanism was stress corrosion.
Evaluation of pipe cracking of the line from the boric acid storage tank to the safety injection system (identified subsequent to the nozzle cracking) revealed that sulfate contamination was the probable cause.
This type of contamination was potentially available to the fuel assembly nozzle because there are pathways to both the spent fuel pool and the primary reactor system from the boric acid storage tank piping.
The potential source of contamination was postulated to be either a contaminated batch of boric acid or resin instru-sion from the recycling system.
Therefore, the primary concerns raised by this j
evaluation are (1) identification of potential j
contaminants, and (2) possible pathways for the contaminants to reach both the spent fuel pool and primary reactor system. A very important aspect of the events is that they appear sympto-matic of a potential generic issue that could involve contamination of the primary reactor system with possible degradation of primary system compo-nents. Previous events involving corrodant attack on stainless steel piping of PWRs were addressed by IE Bulletin 79-17 and IE Circular 76-06.
4 Engineering Date Evaluation Issued Subject E314 6/28/83 LOSS OF ALL THREE CHARGING PUMPS DUE TO EMPTY COMMON REFERENCE LEG IN THE LIQUID LEVEL TRANSDUCERS FOR THE VOLUME CONTROL TANK On October 23, 1982, with St. Lucie Unit 1 in hot standby during recovery from a reactor trip, the three inservice positive displace-ment charging pumps (ARMC0) stopped circulating reactor coolant because the volume control tank (VCT) was pumped dry. Although the VCT was empty, its two liquid level sensors erroneously indicated an acceptable liquid inventory and, hence, an apparently acceptable inflow / outflow balance from the VCT.
The false liquid level indica-1 tion was caused by an empty reference leg that l
was shared by both liquid level sensors. The reference leg was-found to be leaktight and the j
cause of the empty reference leg is not known.
The AE0D staff concluded that the consequences of this event were minor because the charging system is not safety-related at St. Lucie.
However, it is conceivable that this event could be repeated at units with safety-related centrif-ugal charging pumps (some Westinghouse designs utilize these pumps for the high pressure safety injection function) which are prone to gas binding in similar circumstances and at other CE plants similar in design to St. Lucie (e.g.,
Millstone Unit 2) that have taken credit for charging pump injection in their LOCA analysis when they went to stretch power.
Several. reports on the vulnerability of the VCT 4
liquid level interface and of shared fluid coupling of liquid level instrumentation have 1
been included as examples of the potential problems that can occur with this type of '
instrumentation.
(See also Power Reactor Events, Vol. 5, No. 1, pp. 11-12. )
9
_ _ _ _ _ _ _ _ _ _ 2.4 Generic Letters Issued in May - June 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, applicants, and/or equipment suppliers regarding matters of safety, safeguards, or environ-mental significance. During May and June 1983, five letters were issued.*
Generic letters usually either (1) provide information 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 regarding 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 83-19 5/2/83 NEW PROCEDURES FOR PROVIDING PUBLIC NOTICE CONCERNING ISSUANCE OF AMEN 0MENTS TO OPERATING LICENSES Those requirements that directly and significantly affect the way in which the licensee and the NRC staff process operating license amendments was highlighted in this letter to all power reactor and testing facility licensees.
Recipients were asked to review standards for making a "no signif-icant hazards determination" and rule changes for submittals of applications for operating license amendments.
A list was provided of designated state representatives to whom the licensee must give copies of license amendment applications and associated analyses concerning significant hazards considerations pursuant to 10 CFR 50.91(b)(1).
83-20 5/9/83 INTEGRATED SCHEDULING FOR IMPLEMENTATION OF PLANT MODIFICATIONS All operating reactor licensees and holders of construction permits were told that NRC is developing a program to establish realistic schedules for implementation of safety improve-ments, both utility-initiated and NRC-required, at operating reactors. to this generic letter is Amendment No. 91 to the Duane Arnold Energy Center operating license and it represents an approach which is acceptable at present to the NRC.
Generic Letters 83-23 and 83-25 have not yet been issued.
_ Generic Date Letter Issued Subject 83-21 5/11/83 CLARIFICATION OF ACCESS CONTROL PROCEDURES FOR LAW ENFORCEMENT VISITS All licensees of operating nuclear power plants, applicants for operating licenses, and holders of construction permits were reminded that the intent of 10 CFR Part 73.55(d)(1) is to detect unauthorized materials. Applying the regulation to disarm bona fide law enforcement officers within the protected areas is not the intent of that regulation.
83-22 6/3/83 SAFETY EVALUATION OF " EMERGENCY RESPONSE GUIDELINES" All operating reactor licensees, applicants for an operating license, and holders of construction permits for Westinghouse pressurized water reactors were sent a copy of (1) NRC letter of June 1,1983 to Westinghouse Owner's Group (WOG) stating that WOG's proposed Emergency Response Guideline (ERG)
Program is acceptable for implementation and will provide improved guidance for developing emergency operating procedures; and (2) " Safety Evaluation by the Office of Nuclear Reactor Regulation in the Matter of Westinghouse Owner's Group Emergency Response Guidelines," in which NRC states that (a) the guidelines meet the most significant requirements of NUREG-0737 and provide the basis for a significant improvement over current plant emergency operating procedures, and (b) the guidelines contain a number of deficiencies that need to be addressed.
NRC suggests that recipients (1) prepare plant specific procedures which conform to the ERG Program, (2) prepare supplements to the ERGS which cover changes, new equipment, or new knowledge and incorporate these supplements into the procedures, and (3) complete and improve the ERGS to meet long-term NRC requirements, followed by incorporation of improvements into plant specific procedures.
83-24 6/29/83 TMI TASK ACTION PLAN ITEM I.G.1, "SPECIAL LOW POWER TESTING AND TRAINING," RECOMMENDATIONS FOR BWRs All BWR applicants for an operating license and holders of operating licenses for Grand Gulf and LaSalle are required to (1) demonstrate the adverse impact the station blackout (SBO) test
_ _ _ _ _ _ _ _ _ Generic Date Letter Issued Subject will have on their plant equipment, and (2) confirm that the BWR Owner's Group recommendations will constitute compliance with TMI Task Action Plan Item I.G.I.
Since one of the original criteria for I.G.1 special tests (ss stated in the Sequoyah SER) is that the test must not pose a hazard to plant equipment, and because at least three BWR licensees (including the Susquehanna licensee, Pennsylvania Power and Light) have indicated the test poses a hazard to equipment in the drywell, the staff is now recommending that unless the need is identified in the resolution of Generic Issue A-44, the SB0 test should not be required at BWRs.
/
1
_ _ _. 2.5 Operating Reactor Event Memoranda Issued in May - June 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 informatior., 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.
a Date J
OREM Issued Subject 83-01 5/83 GENERIC IMPLICATIONS OF THE MAINE YANKEE FEEDWATER LINE RUPTURE EVENT This memorandum provides a description of the January 25, 1983 loss of main feedwater event at Maine Yankee. The safety significance of the event is the simultaneous occurrence of feedwater line cracking and feedwater system water hammer loads in two of three secondary loops.
Had the event been so severe as to preclude steam generator level recovery in two loops, the third loop would have been available for decay heat removal; however, a two-loop plant in this situation would not be as fortunate.
This memorandum provides a conservative estimate on the likelihood of such an event resulting in a loss of main and auxiliary feedwater in two-loop units.
Applicability 4
to other facilities is also provided, as well as conclusions and recommendations for future investigations and studies of steam generator water hammer.
(See also Power Reactor Events, Vol. 5, No.1, pp. 8-11 for a summary of the Maine Yankee event.)
1
i 2.6 Regulatory and Technical Reports Issued in May - June 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/GP0 Sales Program, PHIL-016, Washington, DC 20555 or on (301) 492-9530.
Report Title NUREG-0020 LICENSED OPERATING REACTORS STATUS
SUMMARY
REPORT (DATA AS Vol. 7, No. 2 0F JANUARY 31,1983)
June 1983 This report provides data on the operation of nuclear units ec timely and accurately as possible.
This information 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-0090 REPORT TO CONGRESS ON ABNORMAL OCCURRENCES (OCTOBER -
Vcl. 5, No. 4 DECEMBER 1982)
May 1983 Section 208 of the Energy Reorganization Act of 1974 identifies an abnormal occurrence as an unscheduled incident 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 October 1 to December 31, 1982. During the report period, there was one abnormal occurrence at the NRC licensees.
The event involved the containment spray system being inoperable at one of the nuclear power plants licensed to operate.
The Agreement States reported no abnormal occurrences to the NRC.
The report also contains information updating some previously reported abnormal occurrences.
l l
NUREG-0304 REGULATORY AND TECHNICAL REPORTS (COMPILATION FOR FIRST l
Vol. 8, No. 1 QUARTER 1983)
May 1983 This compilation lists all NRC regulatory and technical reports published under the NUREG series during the first quarter of 1983.
Report Title NUREG-0485 SYSTEMATIC EVALUATION PROGRAM STATUS
SUMMARY
REPORT Vol. 5, No. 5 (DATA AS OF MAY 31,1983)
June 1983 The Systematic Evaluation Program is intended to examine many safety-related aspects of 11 of the older light water reactors.
This document provides the existing status of the review process including individual topic and overall completion status.
NUREG-0540 TITLE LIST OF DOCUMENTS MADE PUBLICLY AVAILABLE (VOL. 5, Vol. 5, No. 2 N0. 2 COVERS FEBRUARY 1983; VOL. 5, N0. 3 COVERS MARCH 1983)
May 1983; Vol. 5, No. 3 This document is a monthly publication containing descriptions June 1983 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-0606 UNRESOLVED SAFETY ISSUES
SUMMARY
(DATA AS OF MAY 27,1983)
Vol. 5, No. 2 June 1983 Provides an overview of the status of the progress and plants for resolution of the generic tasks addressing " Unresolved Safety Issues" as reported to Congress.
NUREG-0880 SAFETY G0ALS FOR NUCLEAR POWER PLANT OPERATION Rev. 1 May 1983 This report presents and discusses the Nuclear Regulatory Commission's, " Policy Statement on Safety Goals for the Operation of Nuclear Power Plants." The safety goals have been formulated in terms of qualitative goals and quantitative design objectives.
The qualitative goals state that the risk to any individual member of the public from nuclear power plan *. operation should not be a significant contributor to that individual risk of accidental death or injury and that the societal risks should be comparable to or less than those of viable competing technologies.
The quantitative design objectives l
state that the individual and societal risks of nuclear power plant operation should not exceed 0.1% of certain other risks to which members of the U.S. population
_ A subsidiary quantitative design objective are exposed.
is established for the frequency of large scale core melt. The significance of the goals and objectives, their bases and rationale, and the plan to evaluate the goals are provided.
In addition, public comments on the 1982
)
proposed policy statement and responses, to a series of i
questions that accompanied the 1982 statement are summarized.
Report Title NUREG-0927 EVALUATION OF WATER HAMMER EXPERIENCE IN NUCLEAR POWER May 1983 PLANTS This report summarizes key technical findings relevant to the Unresolved Safety Issue A-1, Water Hammer.
These findings were derived from studies of reported water hammer occurrences and underlying causes.and provide key insights into means to minimize or eliminate further water hammer occurrences.
Although these findings provided the technical basis for proposed revisions to NRC's Standard Review Plan, this report does not represent a substitute for current rules and regulations.
NUREG-0986 RCSLK8: REACTOR COOLANT SYSTEM LEAK RATE DETERMINATION June 1983 FOR PWR$ (USER'S GUIDE)
RCSLK8 is a computer program that was developed to analyze the leak tightness of the primary cooling system for any pressurized water reactor.
From system conditions, water levels in tanks, and certain system design parameters, RCSLK8 calculates the loss of water from the cooling system and the increase of water in the leakage collection system during an arbitrary time interval.
The program determines the system leak rates and displays or prints a report of the results. For initial application of the program at a reactor, RCSLK8 creates a file of system parameters and stores it for future use.
RCSLK8 was designed for use in the field with Osborne 1 Portable Computers equipped with double density disk drives.
NUREG-0992 REPORT OF THE COMMITTEE TO REVIEW SAFEGUARDS REQUIREMENTS May 1983 AI POWER REACTORS In October 1982, NRC's Executive Director for Operations appointed a five-member Committee to review NRC security requirements at nuclear power plants with a view toward evaluating the impact of these requirements on operational safety.
During visits to five power reactor sites and more than a dozen meetings over a period of four months, the Committee observed plant operating conditions and obtained views from about 100 persons representing 16 nuclear utilities and industry organizations.
They also interviewed about 40 NRC employees, including Resident Inspectors, and members of the Regional and Headquarters staffs.
Overall, the Committee did not identify any clear operational safety problems associated with implementation of the NRC's i
l security requirements.
However, they did find that the-potential existed, to varying degrees, at licensed facilities.
The Committee's report, dated February 28, 1983, contains five basic findings and a number of associated recommendations intended to minimize the potential impact of security on safety.
J l Report Title l
NUREG/CR-2000 LICENSEE EVENT REPORT COMPILATION (VOL. 2, N0. 4 C0 VERS Vol. 2, No. 4 APRIL 1983; VOL. 2, NO. 5 COVERS MAY 1983)
May 1983; Vol. 2, No. 5 This monthly report contains Licensee Event Report (LER)
June 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.
Pro-cedures for LER reporting are described in detail in the NRC Regulatory Guide 1.16 and NUREG-0161, Instructions for Preparation of Data Entry Sheets for Licensee Event Reports.
The LER summaries in this report are arranged alphabetically by facility name, system, and keyword indexes follow the summa ries.
The components and systems are those identified by the utility when the LER form is initiated; the keywords are assigned by the NOAC staff when the summaries are prepared for computer entry.
NUREG/CR-2044 REACTOR OPERATING EXPERIENCES 1978-1980 May 1983 This compilation contains those experiences reported during the period January 1978 through December 1980 in NRC bulletin Power Reactor Events, which superseded Current Events -
Power Reactors in 1979.
The individual reports were compiled by the Office of Management and Program Analysis with the Office for Analysis and Evaluation of Operational Data assuming this responsibility with the November-December 1980 report (Vol. 3, No.1).
The issues are arranged in chron-ological order.
A reactor plant index and keyword index have been prepared using the Nuclear Operations Analysis Center (NOAC) thesaurus of indexing terms.
In addition, a permuted-title index is included to assist the reader in locating reports of interest.
SUMMARY
REPORT ON A SURVEY OF LIGHT-WATER-REACTOR SAFETY Rev. 1 SYSTEMS May 1983 This report is a revision of NUREG/CR-2069, which described a survey that was performed for the NRC's Division of Risk Analysis.
The purpose of the survey was to collect data on various safety systems for all operating commercial nuclear power plants and review this data for comparison of plants and plant designer's techniques.
Forty-seven plants (69 reactor units) were reviewed.
This report contains an updated and corrected data file and six individual detailed studies of the safety systems.
b 1
l Report Title NUREG/CR-2607 FIRE PROTECTION PROGRAM FOR THE U.S. NUCLEAR REGULATORY June 1983 COMMISSION Since early 1975, Sandia National Laboratories has been conducting fire protection research for the NRC. Testing has been done on grouped electrical cable fires including electrical initiation, fire propagation, the effects of fire retardant coatings and barriers, suppression, and 1
l characterization of the damageability of electrical ca bles.
In addition, several studies of a more generic nature such as fire detection, ventilation, and fire-hazards analysis methodologies were performed.
i This report condenses all of the test results, reports, papers and research findings of the past seven years.
Research conducted by contractors to Sandia is also summari zed.
NUREG/CR-2789 PRESSURE VESSEL THERMAL SHOCK AT U.S. PRESSURIZED-WATER May 1983 REACTORS: EVENTS AND PRECURSORS The U.S. nuclear industry and the NRC recently. (1981-1982) conducted an intensive examination of the capability 4
of pressurized-water reactor (PWR) pressure vessels to withstand severe thermal shocks without compromising their integrity. Many older reactor vessels were manufactured with high copper contents in welds, have accumulated large doses of fast neutrons in service, and have current nil ductility transition temperatures in the 200* to 300'F range.
It is thought that operational temperature transients similar to or more severe than the Rancho Seco incident of March 20, 1978, have the potential to cause certain.preexistent flaws in the wall of these vessels.to propagate.
The potential from the thermal-hydraulic and fracture mechanics viewpoint, but also from the probabilistic viewpoint: What are the probability and consequences of vessel failure due to a specific accident sequence and how can such risks be reduced?
This study is intended to support this examination by i
drawing on the licensee event reports (LERs) of operating:
U.S. PWRs.
The data base is approximately 16,000 PWR LERs stored in the computerized data bank of the Nuclear Operations Analysis Center (NOAC) in Oak Ridge for the j
period 1963-1981.
This data base covers most reportable events at U.S. PWRs, subject to the changes in their reporting requirements over the years.
1 l
l Report Title NUREG/CR-2837 PNL TECHNICAL REVIEW 0F PRESSURIZED THERMAL SH0CK ISSUES May 1983 (TECHNICAL CRITIQUE OF THE NRC NEAR-TERM SCREENING CRITERIA)
Pacific Northwest Laboratory (PNL) provided a technical critique of the draft report, NRC Staff Evaluation of Pressurized Thermal Shock, dated September 13, 1982.
This report provided the basis for the NRC near-term regulatory position on pressurized thermal shock (PTS) and recommended a generic screening criteria for welds in the vessel beltline region.
The PNL staff concluded that the screening criteria were adequate to meet the intent of the NRC safety goal and to retain past predictions of vessel reliability.
The conclusion was based on selecting the plant-specific nil ductility transition reference temperature (RT/NDT) in the conservative manner described within the staff report.
Conservative and unconservative factors were mentioned throughout the NRC staff report.
The PNL staff has listed these factors together with unknown (may be either conservative or unconservative) factors and estimated, where possible, the range in RT(NDT).
The unknown factors were so widespread that the PNL staff recommended that specific conservatisms not be reduced until the unknowns are further resolved.
NUREG/CR-2883 STUDY OF THE VALUE AND IMPACT OF ALTERNATIVE DECAY HEAT Vols. I through 3 REMOVAL CONCEPTS FOR LIGHT WATER REACTORS June 1983 Reliability assessments were made of systems used to remove decay heat from pressurized and boiling water reactors. Current design practices in both U.S. and foreign plants were reviewed in order to identify the types of systems commonly used for.
decay heat removal and to determine the regulatory criteria that control decay heat removal system design in various-countries.
Typical existing decay heat removal system designs were identified for a number of different plant configurations, and the reliability of each of these systems was assessed.
Alternative decay heat removal systems were l
postulated and improverrents in decay heat removal reliability were assessed for various combinations of' plant configurations and alternative decay heat removal concepts.
Alternative concepts that could be implemented by retrofitting existing plants and ones that would be feasible onli for new construction were considered.- Cost estimates were made for those alternatives that provided significant improvements in decay heat removal reliability.
)
NUREG/CR-3047 CLOSE0VT OF IE BULLETIN 80-16:
P0TENTIAL MISAPPLICATION June 1983 0F ROSEMOUNT PRESSURE TRANSMITTERS In March and April 1980, the NRC was. advised by means of.
several 10 CFR Part 21' deficiency reports that a potential.
misapplication problem existed for Rosemount Ins. Models 1151 and 1152 pressure. transmitters with either "A" or
_ Report Title "D" output codes.
It was pointed out that excessive over or reverse pressure could cause ambiguous signals.
In order to bring this potential problem to the attention of licensees and permit holders, and to assure completion of any necessary corrective actions in safety-related systems, the NRC issued IE Bulletin 80-16, June 27,1980. Utility responses for 130 facilities with operating licenses, construction permits or licenses for lower power testing have been evaluated.
The Bulletin has been closed out for 105 (81%) of the 130 current facilities.
Followup items are proposed for 25 current facilities with open status. The Bulletin has been closed out for 5 facilities on the basis of 10 CFR 50.55(e) deficiency reports, which initiate NRC followup on a tracking system which is inde-pendent of this closeout.
From a review of licensee event reports concerning Rosemount components from 1969 through May 1980, it was determined that no event had been caused by the potential problem of ambiguous output from the subject pressure transmitters.
Initially, Type "E" output circuit boards were proposed for replacement in misapplied "A" or "D" output transmitters, but later, samples of the Type "E" boards were found by test to give ambiguous output after exposure to radiation.
Further investigation and testing by Rosemount has resulted in Type "N" replacement output circuit boards for misapplied transmitters which eliminate the possibility of the ambiguous over or reverse pressure signals, with or without degradation from exposure to radiation.
NUREG/CR-3123 CRITERIA FOR SAFETY-RELATED NUCLEAR POWER PLANT OPERATOR June 1983 ACTIONS:
1982 PRESSURIZED WATER REACTOR (PWR) SIMULATOR EXERCISES The actions of 24 teams of nuclear power plant (NPP) operators were recorded by an automatic performance measurement system as they responded to simulated plant casualties during refresher training sessions conducted in the control room simulator for a pressurized water reactor (PWR) plant. The study was part of the Safety-Related Operator Actions program at Oak Ridge National Laboratory, whose purpose is to provide j
a data base to support development of criteria for safety-related actions by NPP operators.
Principal performance measures were initial response times (RTs) and omission errors (0Es), which were defined as the failure to operate specified in a written procedure.
Event RTs appeared to be log-normally distributed.
There was no relation between RT and event severity as indexed by the Plant Process Condition Classifications of the simulated
~
. Title Report casualties. Data on the operator's years of formal education and experience in NPPs was obtained by means of a biographical questionnaire.
There was no relation between the average experience or educational level of l
a team's members and the team's mean rank RT.
The OE rate for process-related controls was 5%.
There was no correlation between RT and the OE rate.
NUREG/CR-3226 STATION BLACKOUT ACCIDENT ANALYSES (PART OF NRC TASK May 1983 ACTION PLAN A-44) l This report presents the results of the accident sequence analysis portion of Task Action Plan A-44, " Station Blackout." Along with a companion report by Oak Ridge National Laboratory on onsite ac power systems, a technical basis for the resolution of Task A-44 is provided.
Probabilistic analysis techniques are used to define and quantify possible accident sequences and identify combinations of failure modes which lead to core melt.
For the dominant sequences, sensitivity analyses are done to show the effects on both individual sequence and total core melt frequency of changing various plant parameters.
Many changes were found to be effective in reducing total core melt frequency.
NUREG/CR-3254 LICENSEE FROGRAMS FOR MAINTAINING OCCUPATIONAL EXPOSURE June 1983 TO RADIATION AS LOW AS IS REASONABLY ACHIEVABLE This report defines the concept of maintaining occupa-tional exposures to radiation as low as is reasonably achievable (ALARA) and describes the elements necessary for specific licensees to implement, operate, and evaluate an effective ALARA program. The rationale for providing more detailed guidance to specific licensees stems from the current recommendations provided by the International Commission on Radiological Protection, as well as from the increased regulatory emphasis on maintaining occupational exposures ALARA.
The objective of this work is to provide the NRC with a basis for updating Regulatory Guide 8.10.
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