ML20081J452

Power Reactor EVENTS.March-April 1983

ML20081J452
Person / Time
Issue date:	10/31/1983
From:	Massaro S, Trenery S NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:
References
NUREG-BR-0051, NUREG-BR-0051-V05-N2, NUREG-BR-51, NUREG-BR-51-V5-N2, NUDOCS 8311080545
Download: ML20081J452 (43)

Text

NUREC/BR-C051

! Q.n POWER REACTOR EVENTS

\\lh..... /

United States Nuclear Regulatory Commission March-April 1983/Vol. 5, No. 2 Power Reactor Events is a bianonthly 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 at1717 H Street. Washington. DC 20555 for a copying fee. Subscriptions and additional or back issues of Power Reactor Events may be requested from the NRC/CPO Sales Program. (301)492-9530. or at PHIL-016, Washington. DC 20555.

Table of Contents Page 1.0 SUMMARIES OF EVENTS 1.1 Unavailability af the Auxiliary Feedwater System.....................................

1 1.2 Personnel injury During Maintenance Activities.................................

4 1.3 Flooding of Reactor Vessel During Refueling................................................

7 1.4 Inadvertent Safety injection Signals During Testing......................................

9 1.5 Improper Use of Lubricant on Valve Seals............

11 1.6 "I" Tube Degradation and Replacement......................................................

13 1.7 Steam Generator Tube Degradation and Repair.............................................

14 1.8 Broken Holddown Springs in Burnable Poison Rod Assemblies......................

14 1.9 Inadvertent Loss of Instrument Air to Salt Water Cooling System..................

16 1.10 References................................................................................................

17 i

i 2.0 ABSTRACTS OF OTHER NRC OPERA TING EXPERIENCE DOCUMENTS 2.1 Abnormal Occurrence Reports (NUREG-0090).............................................

19 2.2 Bulletins and Information No tices.................................................................

20 2.3 Engineering Evaluations and Case Studies....................................................

28 j

2.4 G eneric L e tters...................................................................................

33 2.5 Operating Reactor Even t Memorant.....................................................

36 2.6 Regulatory and Technical Reports (NUREG-0304)......................

37 Editor: Sheryl A. Massaro B311OBO345 831031 Associate Editor: Steven E. Trenery PDR NUREG BR-OOS1 PDR Off. ice for Analys.is and Evaluat. ion of Operational Data U. S. Nuclear Regulatory Commission Published in:

October 1983 Washington, D. C. 20555

1.0 SUMMARIES OF EVENTS 1.1 Unavailability of the Auxiliary Feedwater System On April 19, 1983, the licensec for Turkey Point Units 3 and 4* reported that the auxiliary feedwater ( AFW) system on Unit 3 had been inoperable for possibly up to five days while the plant was operating at 100% power. This i

event was caused by misalignment of two manually-operated valves in the AFW system.

The AFW system is required to be operable during plant startup, shutdown, hot standby, and emergency conditions.

It is an engineered safety features system that supplies the steam generators with a source of feedwater when the main feedwater system is not in service or only small feedwater flows are needed. If it is inoperable, the plant's technical specifications require that the reactor be shut down.

At Turkey Point, the AFW system is shared between Units 3 and 4 (see Figure 1).

Unlike most pressurized water reactors, Turkey Point does not have electrical motor-driven pumps for the AFW system.

Instead, there are three turbine-driven AFW pumps ( A, B, and C), which are common to both units. There is a pair of manually-operated, steam isolation valves in each of the six steam lines, as shown in Figure 1.

These six lines are served from steam headers emanating from the Unit 3 and 4 steam generators such that either unit's steam generators can supply steam to any or all of the three turbine-driven pumps. At the time of the event, Unit 3 was operating at 100% power, Unit 4 cas shut down for refueling and steam generator repair, and the A AFW pump aas out of service for modifications. Under these conditions, only isola-tion valves 3-084A, 3-084B, 3-086A, and 3-086B from the Unit 3 steam line to the B and C AFW pumps were required to be open.

During the time of the event, alternate steam supply lines were being added to each AFW pump turbine. As shown in Figure 1, each alternate line has two manually-operated steam isolation valves in series. Since the alternate system was not yet in operation, all six of the isolation valves (001A, 002A, 001B, 002B, 001C, and 002C) would normally be closed. However, operational testing of the new alternate steam supply lines required extensive tagging and manipulation of these valves.

At approximately 6:05 a.m. on April 19, 1983, a nuclear turbine operator (NTO) reported to the nuclear plant supervisor (NPS) that he had discovered manually-operated valves 3-084A and 3-086B were closed, rendering the AFW system inoperable for automatic operation. A nuclear watch engineer (NWE) &nd the NTO were dispatched from the control room to investigate the status of the AFW system. The NWE confirmed that the two manually-operated valves were closed and, therefore, AFW peps B and C were inoperable. Under the direction j

of the NWE, the two manually-operated valves were opened and system alignment was verified. By 7:35 a.m., the licensee had successfully completed an operability test of the AFW peps. The licensee notified the NRC resident inspector of the incident at 6:30 a.m., and the NRC Emergency Operations Center at 7:00 a.m.

O Turkey Point Units 3 and 4 are each 646 MWe (net) PWRs. They are located 25 miles south of Miami, Florida, and are operated by Florida Power and Light.

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1 FROM FROM FROM FROM FROM FROM UNIT 3 UNIT 4 UNIT 3 UNIT 4 UNIT 3 UNIT 4 3482A 4482A 3484A 4-004A 3406A 4486A X=

k--

X--

X -,,,.

X--

X-,,,.

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+

+

--o-e

--s-e 3483 4483 N

001A 0018 3487 4437 001C 003A 002A 0038 0028 003C 002C i

To"A" "

J To "B" 2

To "C" 2

l AFW Pump' AFW Pump' AFW Pump' Turbine Turbine Turbine

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Figure 1. Turkey Point Units 3 and 4 l

Steam Supply to Auxiliary Feedwater Pump Turbines l

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. Detailed investigations were conducted by both the NRC and the licensee to detemine the cause of the event. Neither investigation could totally reject the possibility that the valve misalignment was due to a willful or inadvertent act; however, the preponderance of evidence suggests that the event was the result of personnel error and that the valves were misaligned (closed) for a period of five days. The following is based on the conclusions reached by both the NRC and the licensee investigations.

On March 26, 1983, a clearance order was issued to close valves 001B, 002B, 001C, and 002C in connection with the ongoing modifications. On April 11, a temporary lift of the clearance order issued on March 26 was issued to permit the opening of 001B and 001C to perfom a hydrostatic test on the new alternate lines. On the morning of April 14, an NTO was given the two clearance tags lifted on April 11, and was instructed to close and retag valves 001B and 001C. The clearance tags referred to valves 001B and 001C, but the tags also referred to the main steam isolation valves to the B and C AFW pumps. On April 19, the NTO reported the apparent valve misalignment. Valve 001B was found closed with its clearance tag removed. This tag was found on valve 3-084A, which was closed. Valve 001C was found open with its tag placed tn valve 3-0868, which was found closed.

The NTO who discovered the misalignment on the morning of April 19, 1983, was the same NTO who had been instructed to close valves 001B and 001C on the morning of April 14. The independent verification of the retagging was required but was never conducted; therefore, the actions taken by the NTO on the morning of April 14 cannot be verified.

Between the mornings of April 14 and 19, 1983, five NTOs made six tours per day through the area of the misaligned valves. The NT0s were required to take pressure and temperature readings and to verify the alignment of the AFW steam supply lines. Apparently, the NT0s failed to identify the improperly closed valves because of inadequately detailed instructions.

The event was considered significant since auxilary feedwater flow is expected to initiate automatically upon loss of nomal feedwater flow.

If nomal feed-water flow is interrupted without initiation of auxiliary feedwater flow, proper operator actions become crucial to ensure that the core is not damaged.

It should be noted, however, that the licensee had detailed operating procedures for dealing with malfunctions of the AFW system, including complete failure.

Westinghouse (the nuclear steam supply system vendor) has performed generic analyses

• which examine the loss of both the main and AFW systems. They conclude that timely operator actions at most plants (i.e., depressurizing the primary system through PORY operation and supplying makeup water via the safety injection system) could prevent core uncovery by expanding the time available to restore feedwater cooling flow.

WCAP-9600, Report on Small Break Accidents for Westinghouse NSSS Systems, i

and WCAP-9744, Loss of Feedwater Induced Loss of Coolant Accident Analysis Report.

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Both the NRC and the licensee concluded from independent investigations l

that personnel error appeared to be the probable cause of the event.

In i

closing the wrong valves and rehanging the tags intended for valves 001B and 1

001C, the NTO apparently misunderstood which valves were supposed to be l

closed.

Instead of closing valves 001B and 001C, the NTO apparently closed valves 3-084A and 0-086B after unlocking them and tagged them accordingly.

Corrective actions taken or planned by the licensee are being monitored by the NRC. The licensee has verified all accessible Unit 3 safety-related flow paths, instrumentation, and main electrical alignments. The B and C AFW pianps are being tested daily until all-major construction work on the AFW pumps and piping is i

compl eted. Written instructions have been issued to operators emphasizing the need for independent verification, the importance of auxiliary feedwater opera-i tions activities, and the importance of performing assigned jobs with accuracy and completeness. The NTO logsheets have been revised to improve detail and clarity. Identification tags were immediately placed on all valves that were previously unlabeled.

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In addition, a new requirement involving monthly walkdowns of all accessible safety system flowpaths will be fonnalized in the plant's technical specifica-tions. Procedures will be revised to emphasize the requirement for independent verification of actions taken when clearance orders are tempararily lifted, I

including the checking of associated tags.

)

The NRC inspected the plant on April 19-28, 1983 with respect to the corrective actions taken by the licensee, and will continue to monitor this event. An enforcement conference was held on April 28, 1983 between licensee and NRC 1

Region II personnel. On August 15, 1983, the NRC sent to the licensee a I

notice of violation and proposed civil penalty of $100,000. (Refs.1 through j

4.)

l 1.2 Personnel Injury During Maintenance Activities The following three events involving industrial type accidents resulting in serious personnel injury occurred during March and April 1983, and were reported to the NRC Emergency Operations Center. Although the NRC does not have regulatory responsibility for other than the radiological aspects, the i

events are included for their general interest and the lessons apparent to prevent recurrence.

t i

Ginna*

On March 31,1983 at 6:15 a.m., a health physics technician working alone l

entered reactor containment to install a radiation monitoring device on the refueling bridge manipulator crane. While working, he fell a distance of about 50 feet from the manipulator crane onto a fuel upender in the empty l

reactor cavity. He was found at about 6:56 by a fire watch, who reported

r. medical emergency to the control room.

Ginna is a 470 MWe (net) PWR located 15 miles northeast of Rochester, New York, and is operated by Rochester Gas and Electric.

Within ten minutes, a medical team entered primary containment. The technician was placed on a "back board" and moved to a Stokes Basket stretcher which subsequently was raised to the refueling floor. An ambulance arrived on site

.at about 7:24 a.m.

The technician's protective clothing was removed, he was transferred to the ambulance via the equipment hatch, and was taken to a local hospital.

En route to the hospital, the technician was surveyed for radioactive contamina-tion with a survey meter equipped with a thin window GM tube.

He was found to have minor contamination confined to small, localized areas, and was decontaminated at the hospital.

Injuries included a collapsed lung, broken ribs, a broken leg, internal bleeding, and spinal injury.

On April 1,1983, an NRC inspector toured the refueling floor of the primary containment and entered the area of the reactor cavity where the injured technician was found. During the tour, the inspector noted several matters of an apparent industrial safety concern. The following matters were discussed with representatives of the Occupational Safety and Health Administration (OSHA):

(1) two step-off pads, used for contamination control, were located close to the east edge of the reactor cavity (see Figure 2);

(2) a single length of rope was being used as a railing at the northeast edge of the reactor cavity, and was attached to the railing at the north end of the cavity and the top of the removable cavity stairwell (Figure 2);

(3) the southeast edge of the reactor cavity did not have railing (Figure 2).

The inspector discussed the above with licensee representatives, who infomed him that a committee had been established to review the event and determine any short and long tem corrective actions.

In addition, OSHA representatives also were at the site to review the event.

(Refs. 5 through 7.)

Surry Unit 1*

On April 1,1983, a Westinghouse employee was injured when struck on the head by a loose section of metal tubing that fell from the reactor coolant ptsnp oil collection system line on which he was perfoming maintenance. The falling object knocked a 6-in gash in the back of his head and knocked him unconscious.

Licensee personnel administered first aid, and the worker regained consciousness before being sent to a hospital.

The licensee detemined that the worker had contamination amounting to 4000 to 5000 disintegrations per minute (dpm) around the wound, about 40,000 dpm on one ear, and 120,000 dpm on the skin on a spot on his chest.

No effort was made to remove the contamination due to the possible serious-ness of the injury. However, first aid was administered, the worker's protective clothing was removed, and he was transported in the licensee's station anbulance to the nearest hospital having the facilities to receive and treat patients bearing radioactive contamination.

Surry Unit 1 is a 775 MWe (net) PWR located 17 miles northeast of Newport News, Virginia, and is operated by Virginia Electric and Power.

' 1 Steam Generator

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z 1A l

4 o

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Head Storage Reactor j

OPad F

Steam j

3 Generator 8

7

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Location of Individual (bottom of reactor cavity, on fuel upender)

2. Personnel Access Hatch

3. Equipment Hatch

4. Rope at Northeast Cavity Edge

5. Step-off Pad (top of verticalladder)

6. Stepeff Pad (top of ladder)

7. Location of Refuel Bridge

8. No Railing at this Ledge e

i Figure 2.

Reactor Contairiment Operating Floor Plan l

l 1

1 The licensee reported that approved procedures for treating, transporting, and admitting contaminated personnel were followed throughout the event.

(Refs. 8 and 9.)

Crystal River Unit 3*

A licensee employee collapsed at 11:30 a.m. on March 23, 1983, while working in the plant's low-level waste handling building, and died about two hours later at a nearby hospital.

The 24-year-old employee was wearing cloth protective clothing with an air-supplied hood, and complained of fatigue before being stricken.

It was detemined that the air supply system hcd been functioning properly, and that the victim had received no rcdioactive contamination.

Review of the employee's records showed that he had undergone medical certifica-tion. An autopsy perfomed on the victim revaaled that he had an enlarged heart; the cause of death was undetemined.

v The plant physician indicated that employees receive a medical review and a pulmonary function test to detemine if an individual qualifies as a user of a i

respiratory protective device, though EKGs are only included for full time empl oyees. The physician indicated that more individuals seem not to qualify for cardiovascular reasons than pulmonary function problems.

(Refs.10 and 11.)

1.3 Flooding of Reactor Vessel During Refueling During February and March 1983, the following similar events occurred at the Browns Ferry and Peach Bottom facilities.

Browns Ferry On February 16, 1983, during a refueling outage, preparations for a containment integrated leak rate test were being perfomed on Unit 2 with drywell pressure at 50 psig.** The vessel head was in place with the head fastening nuts not installed and the head vent open as required by test procedures.

Apparently due to low water level and high drywell pressure signals, four core spray peps, four residual heat removal (RHR) pups, and eight diesel generators started. The RHR system was secured before injection into the vessel occurred. A total of 44,000 gallons of water were injected into the vessel from the torus via the core spray system, which caused spillage into the drywell smps and put some water into the steam lines. (The spillage was from the head vent and from the momentary lifting of the reactor vessel head caused by the pressure buildup form the core spray pap discharge.) The torus water level dropped 6-1/2 inches. Water chloride and conductivity remained within technical specification limits.

Crystal River Unit 3 is a 782 MWe (net) PWR located seven miles northwest of Crystal River, Florida, and is operated by Florida Power Corporation.

Browns Ferry Unit 2 is a 1065 MWe (net) BWR located ten miles northwest of Decatur, Alabama, and is operated by Tennessee Valley Authority.

/

All injection pumps were secured. Water levels were returned to nomal in both the reactor vessel and the torus, and the steam lines were drained.

The cause of the spurious initiation signals remains under investigation by the licensee.

(Refs. 12 and 13.)

Peach Bottom On March 3,1983, during the Unit 3* refueling outage, an inadvertent initiation of the 3A and 3B RHR pups transferred approximately 65,000 gallons of water from the torus into the reactor. The initiation was caused by a false low reactor water level signal which was present for less than 3.5 seconds. The protective logic responded as if a true low level signal was present. Therefore, the automatic actions associated with a true loss-of-coolant accident (LOCA) signal occurred. Accordingly, the operating 3B recirculation pump tripped, the El and E4 diesel generators started, the 3A and 3B RHR pumps started and the injection valves opened, the recirculation purnp discharge valves closed, and the high pressure service water pumps tripped. At the time of the occurrence, the E-2 diesel generator was already running, for reliability reasons, with the E-3 diesel generator blocked for maintenance.

Because the initiating low level signal cleared within 5 seconds, the LOCA logic sequence did not continue.

Since the unit was in refueling, with the reactor cavity flooded, most of the water overflowed into the cavity ventilation ducting, onto the refueling floor and down the main hatchway to the 135-ft elevation, where approximately 50 gallons flowed out of the building and into the storm drain system. The water-filled ventilation ducting ruptured in several places and water cascaded down all levels of the reactor building.

When the water from the reactor cavity and spent fuel pool began flooding the refueling floor, personnel in the Unit 3 reactor building and drywell exited the area. The 3A and 3B RHR pumps were tripped within approximately 4 minutes and the RHR injection valves were closed. The torus header valve and full flow test valve were opened, along with the manual discharge valve to the condensate storage tank from the reactor water cleanup system letdown, to restore reactor water level to within shutdown range.

Immediate actions were taken to assess potential leakage paths to the environ-ment. The water flow to the drain system at the 135-ft elevation was halted.

Sampling of the water at various locations along the storm drain flow path was begun. Water samples taken from the storm drain system along the discharge route to the Susquehanna River indicated decreasing activity. Samples taken at the discharge to the river indicated no detectable activity due to the I

dilution of the small amount of radioactive water by the normal' flow in the storm drain system. Clean-up and decontamination were initiated in the Unit 3 reactor building and refueling floor.

Peach Bottom Unit 3 is a 1035 MWe (net) BWR located 19 miles south of Lancaster, Pennsylvania, and is operated by Philadelphia Electric.

l l

The sensed low level which initiated the LOCA signal was caused by instrianent technicians venting nitrogen from a pressure switch into the reference leg of the 2B Yarway instrumentation loop upon completion of a surveillance test.

Af ter the occurrence, the E-2 diesel generator LOCA signal, which can only be reset by removing the diesel from service, remained sealed in. This signal also removes the diesel's Cardox system automatic initiation feature. A fire watch was posted when this fact was recognized until the LOCA signal reset was achieved. (Refs. 14 through 16.)

1.4 Inadvertent Safety Injection Signals During Testing Summer Unit 1* experienced inadvertent safety injection signals on March 18 and 19, 1983, during reactor cooldown and testing of reactor trip breakers (RTBs). The testing was being perfomed in accordance with NRC Inspection and Enforcement Bulletin (IEB) 83-04, " Failure of Undervoltage Trip Function of Reactor Trip Breakers," which was issued on March 11,1983 (see p. 21 for an abstract of this bulletin).

The first event occurred on March 18, 1983, with the plant in hot shutdown.

Reactor coolant system (RCS) pressure was approximately 1100 psig and temperature was approximately 345'F. Maintenance personnel were in the process of performing reactor trip breaker testing for train A, as required by IEB 83-04.

The technician responsible for performing the test had not been adequately briefed and was under the assumption that he was to perfom the entire solid state protection system surveillance procedure. He had previously perfomed this test and was familiar with the normal method of testing.

When the technician requested release to work from the control room foreman, the foreman told him that only a partial test was to be performed and that the blocks associated with the low pressurizer pressure and steam line safety injection were not to be unblocked. The technician reviewed the procedure and found the step associated with the RTBs.

(This is part of the Semi-automatic Test Circuitry, but is addressed as a separate step.)

In reviewing this step, the technician found no reason why he could not meet the criteria that the control room foreman had specified. Pemission was given to perfom the testing based on the technician's knowledge. The technician perfomed the precautions and initial conditions sections of the procedure and proceeded to the step fu RTB testing. This procedure requires that the switches on the test panel be positioned to preselected positions. The technician then selects any one of the four nomal input function test pushbuttons.

In this case, the technician depressed a function switch (No.1), which causes the output relays for train A safety injection to " pickup," and the train A

^

RTBs to trip. Thus, a safety injection was idtiated.

(It should be noted I

that if another function switch (No. 3) had been depressed, the test would have been completed without a safety injection at tuation.)

Reactor coolant i

system wide range pressure recorders indicated that the RCS pressure increased to approximately 1550 psig during the event. However, control room personnel stated that they observed RCS pressure reaching 2000 psig, at which time the safety injection signal was blocked and the safety injection was terminated.

i Summer Unit 1 is a 900 MWe (net) PWR located 26 miles northwest of Columbia, South Carolina, and is operated by South Carolina Electric and Gas, Subsequent testing perfomed on the subject recorders indicated that they were perfoming properly, thereby supporting the case that RCS pressure peaked at approximately 1550 psig. However, assuming that the operator observations were correct, the condition for exceeding 1600 psid on the steam generator support This evaluation indicates plate and tubes has been evaluated by the licensee.

that the inadvertent safety injection produced negligible impact on the integrity of the steam generators.

The cause of the inadvertent injection has been attributed to personnel Corrective action has been taken in order to prevent recurrence of error.

the conditions which initiated this safety injection.

The technician involved has been counselled to understand that any deviations from normal routineThe (procedures) should be reviewed by supervisors prior to implementation.

i maintenance foreman also has been counselled on the fact that the technician should be briefed fully of the job to be perfomed.

If a procedure is to be perfomed, in part, a thorough evaluation or review of tae procedure and reference manuals should be done prior to perfoming the test.

Also, during subsequent analysis of this event, it was recognized that the licensee's Emergency Operating Procedure (EOP) based upon the Westinghouse

)

Technical Guidelines, constrained the operators from terminating safety At the time J

injection flow prior to acnieving RCS pressure of 2000 psig.

of the safety injection, the primary system temperature was 340*F and pressure was 1000 psig while the main steam pressure was 100 psig.

Consistent with their procedures, the reactor operators did not terminate safety injection flow until the RCS pressure reached 2000 psig. This generated a pressure differential across the steam generator tubes which exceeded the 1600 psi design dif ferential pressure by about 300 psi.

The problem arose because the generic technical guidelines, which include a 2000 psig constraint, were not intended for application during reactor cooldown, and the plant procedures did not differentiate for this condition.

18, 1983, a station order was issued to the operations group to ensure On March that operations personnel were cognizant of this fact and aware of the fact that they have the authority and responsibility to take actions necessary to protect the facility and the health and safety of the general public.

This condition will be further evaluated in order to detemine if procedure and/or system changes are necessary.

A second inadvertent safety injection occurred at Summer Unit 1 on March 19, 1983, with the plant in cold shutdown. The RCS was water solid, at approx-imately 325 psig and 150*F. Maintenance personnel were in the process While returning of performing RTB testing for. train B, as required by IEB 83-04.

the solid state protection system to service following the required RTB testing, the manual blocks for pressurizer low pressure safety injection and steam ifne low pressure safety injection were reinstated by operating the block reset switches on the main control board. A few seconds after completion of this step of the procedure, a "first out" alarm was received for low pressurizer pressure safety injection. The event was immediately recognized as being spurious and the reactor operator teminated the safety injection after ipproximately two minutes without waiting for the E0P tennination criteria of 2000 psig RCS pressure to be reached. The RCS was water-solid, the plant's cold overpressure protection system was in service and the power-operated relief valve (PORV) was set to relieve at about 750 psig. However, because of the prompt operator action, safety injection was terminated at approximately 550 psig before the PORV setpoint was reached.

In this instance, because of the plant s operating mode, terminating prior to meeting the E0P termination criterion was appropriate.

The cause of the inadverb at safety injection has been attributed to malfunction of the associated block / reset switches on the main control board, resulting in the pressurizer low pressure safety injection becoming reset for train A.

Each switch consists of the " block" and " reset" positions with a spring return to the center position.

To prevent recurrence of this condition, an awareness training instruction has been issued to the operations personnel alerting them of this occurrence.

It cautions the operators to operate these switches with care, and to verify the status indication for the block / reset fu'iction. This infonnation also has been included in the operator requalification training program. In addition, an evaluation will be performed to detennine if the main control board switches can be " separated," so that the block and reset functions will not be accomplished with the same switch.

The occurrence of the above events resulted in the issuance of IE Information Notice No. 83-30, " Misapplication of Generic Emergency Operating Procedures (E0P) Guidelines," i.. May, 1983. This infonnation notice was provided as notification of the potential misapplication of generic E0P technical guidelines to operating modes for which they were not designed and to conditions for which they do not apply. (Refs.17 through 20.)

1.5 Improper Use of Lubricant on Valve Seals On April 7,1983, during a scheduled refueling outage at Dresden Unit 2,*

the NRC Resident Inspector received an anonymous phone call from an individual implying that grease may have been used on the seating surface of main steam isolation valves (MSIVs) to aid in the passing of local leak rate tests (LLRTs).

Based on this inquiry, an NRC investigator and inspector were dispatched to the site where mechanical maintenance personnel inspected the seating surfaces l

of three MSIVs that had recently been repaired and successfully local leak rate tested. The valves were disassembled, inspected, and Dow Corning 111 grease (silicone) was found on the disk and seat of all three.

(The vendor maintenance manual for these valves requires valve disks and seats to be clear of grease and foreign substances.) After cleaning and reassembly, only one valve passed the LLRTs. The two remaining valves were disassembled again, the seats relapped, and the valves reassembled. When all maintenance was y

completed and the valves assembled, both valves passed the LLRTs.

1

)

Dresden Unit 2 is a 772 MWe (net) BWR located nine miles east of Morris, j

Illinois, and is operated by Commonwealth Edison.

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l

. After grease had been discovered on the seating surface of these MSIVs, a list was prepared of those valves in which the valve seat was exposed during this outagt ard could have been greased during maintenance or repair. Twenty valves foming part of the primary containment isolation boundary, including the MSIVs, were identified. Af ter the list had been studied, it was detemined that there was reason to believe that three other valves had greased seats.

Additionally, a sample of three of the remaining valves believed to be free of grease were selected for inspection to verify their condition.

These six additional valves were disassembled and inspected. Grease was found on three of them. The seating surfaces of these three valves were cleaned, and all six valves were reassembled. Five valves successfully passed the LLRTs without further maintenance. The sixth valve was dis-assembled once again, the "0" ring seating surface between the valve seat assembly and the valve body was relapped, and the "0" ring replaced. When the valve was reassembled, it was successfully local leak rate tested.

On April 11, 1983, the licensee and NRC regional staff agreed that the following corrective actions will be taken:

(1) A comprehensive review of other isolation valves maintained during the current Unit 2 and the previous Unit 3 refueling outage, to address generic ccncerns of inappropriate use of grease during valve reassembly.

(2) A review of the event by the licensee's Quality Control Department, and development of standard inspection plans for isolation valve maintenance requiring valve disassembly and reassembly.

I (3) Preparation and presentation of a review of the event and its implications to mechanical mintenance supervisors, to emphasize the critical nature of the decision to use lubricants where not specifically addressed in the procedure.

l (4) Review and appropriate revision of existing maintenance procedures dealing with assembly and disassembly of valves, to ensure appropriate precautions and controls are in place with regard to the use of lubricants and sealants.

1 (5) Preparation and incorporation of a training module into the maintenance training program, dealing with the correct use of lubricants and i

sealants in various maintenance activities.

l (6) Development of a repair manual, specifically for disassembly and reas-l sembly of MS1Ys.

In addition to the commitment to complete the action items listed above, the licensee perfomed en internal audit of the incident.

A committee was fomed to investigate the mintenance practice of using grease during the reassembly of valves subject to post-maintenance LLRTs. The goal of the committee was to investigate the particular circumstances involved in the use of grease during the reassembly of MSIVs, and also to address the broader question of the implication of the maintenance practice on other critical valves at Dresden. As a result of the investigation, it was detemined by the licensee that the motivation behind the use of lubricants and/or grease in the reassembly l

of the MSIVs and other valves was not to intentially bias the results of any

i LLRT, but to assist in the reassembly of large and sometimes difficult to maintain equipment. The NRC concurred with this conclusion. Tracking of the i

six action items will be done to ensure follow-up and completion.

(Refs. 21 through 23.)

1.6 "J" Tube Degradation and Replacement In early 1982 at the Ginna facility, inspections were made at the request of Westinghouse (the nuclear steam system supplier) of the secondary side of a steam generator. These inspections identified apparent surface erosion on i

the inside of the 38 "J" tubes on the feedwater ring (i.e., the feedwater supply distribution header within the steam generator). Wall thinnir.g was l

initially noted visually on the inner surface of the extrados side of the i

bend of the outlet of the "J" tube. Subsequently, ultrasonic (UT) wall

{

thickness measurements confirmed the wall thinning and indicated it to be throughout the length of the tube. These results were reported to Westinghouse in April 1982.

In November 1982, Westinghouse responded with information on the status of the "J" tube erosion and recommended replacement with Inconel Alloy 600 material. The response discussed the ASME code require-ments (the "J" tube to feedwater ring weld is a non-code internal vessel-weld) and other reported experience which indicated that erosion can lead to t

loose parts, a potential safety concern. The "J" tubes, themselves, were

]

a modification added to the top of the feedwater ring as part of the design change to reduce the potential for steam generator feedwater water hammer; the outlet holes in the bottom of the feedwater ring were plugged at the same time.

i Westinghouse (W) promulgated an advisory notice to 12 domestic PWR plants concerning the wall thinning of "J" tubes in (W) Technical Bulletin NSD-TB--

i 82-07 dated December 28, 1982.

In this bulletTn, W indicated that "J" tube i

operation with 50% of the nominal 0.218" wall thicEness is acceptable. W i

estimated (on a linited data basis) that erosion rates were in the order of 0.030 inches per year. The W recommended action for the plants was inspection i

of "J" tubes in at least one steam generator that had three years service with "J" tubes and replacement of carbon steel "J" tubes with less than 50% wall thickness with Inconel Alloy 600.

In April 1983, the licensee again performed UT examinations of the "J" tubes on the steam generator feedwater ring header and found significant metal loss from the "J" tube walls. Nominal wall losses of up to 50% were found in the majority of the tubes and holes were found in about a dozen "J" tubes near the attachment weld and in the neck area. The root causes of the erosion (and possibly corrosion) problems are under review and involve materials, water velocity, water chemistry, and other factors. An inspection of the lower area of the feedwater ring did not indicate any leaking areas (during an earlier j

outage, inspection revealed minor leaks around some welded plugs which were repaired).

l The licensee modified the NSSS recommendations by machining the existing "J" the welds to produce a flat " spot-face" weld preparation rather than to return the joint to the saddle fit geometry. This modification was a considerable improvement in joint preparation.

It was accomplished with special portable machining tools.

i 4

The licensee replaced all the "J" tubes prior to restart. The generic l

implications of these problems is being reviewed. (Ref. 24.)

)

1.7 Steam Generator Tube Degradation and Repair In April 1983, a program of p1anned eddy current examination was conducted i

on both steam generators during a refueling and maintenance outage at Ginna.*

The examination consisted of multifrequency eddy current examination to i

detect and measure potential tube defects. The inlets of both steam generators received 100% inspection and the outlets of both steam generators received approximately 25% inspection. Previous inspections had identified tube sheet crevice intergranular attack (IGA) as causing tube degradation.

The licensee found indications on 78 tubes in the B steam generator inlet and four tubes in the A steam generator inlet, but no indications on the outlets of either steam generator. Of these, 23 tubes in the B inlet were identified as having defects greater than inservice inspection acceptance criteria (>40%).

Sixteen tubes had previous indications, while seven were new indications. An additional 55 tubes gave indications that were below the plugging criteria l

(36 were new ones). The licensee concluded that concentrations of caustics and/or alkaline earths in the tube sheet crevice led to an electro-chemical intergranular attack (IGA) of the tubing. A program of crevice flushing has been conducted a number of times since 1980 and is continuing. These flushes renove active species and contaminants from the crevices, reducing the severity of attack to the tubing.

The tube repair techniques consisted of plugging or of inserting one of three varying length sleeves in the tubes with IGA indications. The sleeve length (22-in, 28-in, or 36-in) chosen for the insertion in a given tube was determined by the vertical room available in the hemispherical channel head. The 22-in sleeves were explosively welded to the tube at both ends while the longer length sleeves were brazed to the tube at the upper end and explosively welded at the lower tubesheet end. In total, 76 sleeves were installed in the B steam generator and four were installed in the A steam generator. Also, two tubes were plugged in the B steam generator.

Problems were encountered with the quality of the initial brazes as evidenced by ultrasonic inspection. Licensee investigation concluded that the tubes were apparently " locked up" between the tubesheet and first support plate, preventing adequate thennal expansion of the tubes during the brazing process.

The sleeving method was revised to provide pretensioning of the tube below the braze location to relieve thennal stresses during brazing. Subsequent brazing l

operations provided acceptable results.

(Refs. 25 through 27.)

1.8 Broken Holddown Springs in Burnable Poison Rod Assemblies On March 10, 1983, during inspection of the fuel assemblies at McGuire Unit 1,**

the holddown spring for a non-fuel bearing component (NFBC) was observed to be Ginna is a 470 MWe (net) PWR located 15 miles northeast of Rochester, New York, and is operated by Rochester Gas and Electric.

McGuire Unit 1 is a 1180 MWe (net) PWR located 17 miles north of Charlotte, l

North Carolina, and is operated by Duke Power.

l' li -

broken. This discovery led to the inspection of all NFBC holddown assemblies for the Unit 1 core. Of the 94 NFBCs equipped with the same spring design, 21 were detemined to be broken. Of these 21 broken springs, three were identified as having double fractures such that a semicircular spring section might move free of the yoke guide and into the flow of the coolant system.

The purpose of these springs is to hold NFBCs (which include thimble plugs, burnable poison rods and sources) in the fuel assemblies, resisting the lifting force of reactor coolant flow. Prior to McGuire Unit 1 initial fuel loading in 1981, Westinghouse (the plant's nuclear steam supply system vendor) had identified deficiencies in their design. Experience at other plants had shown that springs identical to those at McGuire were prone to breakage in the first 3000 hours0.0347 days <br />0.833 hours <br />0.00496 weeks <br />0.00114 months <br /> of operation. The close match of spring and reactor coolant punp resonances was theorized to be a contributing factor to the failures, as was poor spring heat-treating during manufacture. Ar.other consideration was the barrel-shape of the spring, which placed the center coils further into the coolant flow path.

At that time, analyses indicated that single location fractures were the most likely failure mode within the first fuel cycle. This was not considered a safety concern.

It was decided that old design springs in the NFBC holddown assemblies which were intended to remain in the core past the first cycle (thimble plugs and secondary sources) would be replaced with a new design.

This new design holddown spring has a cylindrical outline and is tuned to a higher frequency. It was installed in the holddown assemblies of the 44 thimble plugs and two secondary sources. This modification (exchanging spring designs) had been completed by October 1980. Thus, the initial Unit 1 core load consisted of 94 NFBC holddown assemblies equipped with the old spring design, and 46 holddown assemblies equipped with the new spring design. After discovering the broken springs in the holddown assemblies of 21 NFBCs in March 1983, an inspection was perfomed on the 46 new design springs used in the holddown assemblies of the thimble plugs and secondary sources. All of these springs wre confirmed to be intact and undamaged.

Upon notification of the broken springs, both the licensee and Westinghouse began evaluating alternate core designs which minimized or eliminated use of the defective holddown springs. A core design has been developed which reloads all fuel assemblies in their previous locations, replaces all burnable poison assemblies with thimble plugs, inserts two new secondary source assemblies, and reinserts the two secondary sources ;each of which contain 16 partially depleted burnable poison rodlets). Therefore, all springs of the defective design have been removed from the core.

By replacing NFBCs containing old design holddown springs with thimble plug assemblies incorporating new design springs, the licensee believes the likelihood of spring failure will be significantly reduced.

The cause of this event is design deficiency, since the broken holddown springs are theorized to be the result of fatigue failure. Three of the damaged springs were confimed to have double breaks, but all pieces were

- _ _ _ - _ _ _ _ _ _ _ l 1

l No retained around the hub of the handling portion of the core component.

other damage or indications of adverse wear were observed. Accurate assess-ment of the broken spring parts remains under review by Westinghouse and the licensee.

(Ref. 28.)

1.9 Inadvertent Loss of Instrument Air to Salt Water Cooling System At San Onofre Unit 1* on March 16, 1983, with the unit in cold shutdown, a construction worker drilling holes in the circulating water intake structure accidently drilled through an instruent air line which supplies the two salt water cooling pep discharge valves. The rupture of this line made it impossible to operate the closed north salt water cooling pep discharge However, the south salt water cooling pep was running at the time, valve.

and its discharge valve remained open following the loss of instrment air.

The salt water cooling system supplies cooling water to safety equipment such as heat exchangers and necessary auxiliary safety-related equipment.

It is required for nomal operation and safe shutdown of the plant under certain transient conditions. During this event, the ability to operate the open south salt water cooling pep discharge valve was considered to be uncertain. The licensee therefore repcrted this event as a significant event because of the apparent potential for total loss of the salt water cooling system if the running south pmp failed.

The north Temporary repairs were completed nine hours af ter the event started.

p m p discharge valve was opened, and the p op started. The licensee kept both pumps running until pemanent repairs to the instrment air line were completed.

(Ref 29.)

\\

San Onofre Unit 1 is a 436 MWe (net) PWR located five miles south of San Clemente, California, and is operated by Southern California Edison.

- 1.10 References (1.1) 1.

NRC, Preliminary Notification PNO-II-83-34, May 5,1983.

2.

Florida Power and Light Company, Docket No. 50-250, Licensee Event Report 83-07, May 3,1983.

3.

Letter from J. O'Reilly, NRC/ Region II, to E. Adomat, Florida Power and Light Company, transmitting a Notice of Violation and Proposed Imposition of Civil Penalities, Docket No. 50-250, August 15, 1983.

4.

Letter from D. Nussbauner, NRC, to all Agreement States, March 25, 1983.

(1.2) 5.

NRC, Preliminary Notification PNO-I-6 -24, March ^',1983.

6.

Letter from B. A. Snow, Rochester Gas and Electric Crmpany, to NRC/ Region I, March 31, 1983.

7.

NRC/ Region I, Inspection Report No. 50-244/83-06, May 13, 1983.

8.

NRC, Preliminary Notification PN0-II-83-19, April 1,1983.

9.

NRC/ Region II, Inspection Report No. 50-280/83-11, April 29, 1983.

10. NRC, Preliminary Notification PNO-II-83-17, March 23,1983.

11. NRC/ Region II, Inspection Report No. 50-302/83-15, May 26, 1983.

(1.3) 12. NRC, Preliminary Notification PN0-II-83-10, February 17, 1983.

13. NRC/ Region II, Inspection Paport No. 50-260/83-05, March 16, 1983.

14. NRC, Preliminary Notification PN0-I-83-15, March 4,1983.

15. Philadelphia Electric Company, Docket NO. 50-278, Licensee Event Report 83-07, March 17,1983.

16. NRC/ Region I, Inspection Report No. 50-277/83-05, April 8, 1983.

(1.4) 17. South Carolina Electric and Gas Company, Docket No. 50-395, Licensee l

Event Report 83-27, April 15,1983.

18. Letter from 0. W. Dix.on, Sr., South Carolina Electric and Gas Company, to J. P. O'Reilly, NRC/ Region II, re:

Inadvertent Safety injections at Virgil C. Summer Nuclear Station, April 15, 1983.

19. NRC, AE0D Technical Review Report No. T320, Inadvertent Safety Irtiections Attributed to Personnel Error at the Virgil C. Summer Nuclear Station, June 14, 1983.

l l

20. NRC, Information Notice 83-11, May 1983.

t j

(1.5) 21. NRC, Preliminary Notification PNO-III-83-26, April 8,1983.

22. Commonwealth Edison Company, Docket NO. 50-237, Licensee Event i

Report 83-27, May 5,1983.

23. NRC, Inspection Report No. 50-237/83-11, July 5,1983.

i (1.6) 24. NRC/ Region I, Inspection Report No. 50-244/83-12, June 17, 1983.

i l

(1.7) 25. Rochester Gas and Electric, Docket No. 50.244, Licensee Event Report 83-13, May 5,1983.

1

26. NRC/ Region I, Inspection Report No. 50-244/83-10', May 25,1983.

{

27. NRC/ Region I, Inspection Report No. 50-244/83-12, June 17, 1983.

(1.8) 28. Duke Power Company, Docket No. 50-369, Reportable Occurrence No.

369/83-11, March 24,1983.

(1.9) 29. NRC, Preliminary Notification PN0-V-83-14, March 17,1983.

These referenced documents are available in the NRC Public Document Room at I'

1717 H Street, Washington, D.C.

for inspection and/or copying for a fee.

I i

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- 2.0 ABSTRACTS OF OTHER NRC OPERATING EXPERIENCE DOCUMENTS 2.1 Abnormal Occurrence Reports (NUREG-0090) Issued in March-April 1983 An abnormal occurrence is defined in Section 208 of the Energy Reorgan-ization Act of 1974 as an unscheduled incident or event which the NRC determines is significant from the standpoint of public health or safety.

Under the 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 NUREG-0090 series of documents. Al so included in the quarterly reports are updates of some previously reported abnormal occurrences, and summaries of certain events that may be per-ceived by the public as significant but do not meet the Section 208 abnormal occurrence criteria.

No reports to Congress on abnormal occurrences were published during this report period.

______- 2.2 Bull,etins and Information Notices Issued in March -April 1983 The Office of Inspection and Enforcement periodically issues bulletins and infomation notices to licensees and holders of construction pemits.

During the period, three bulletins and 19 infomation 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 prcblem, 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 infomation NRC should have to assess the need for further actions.

A prompt response by affected 'icensees is required and failure to respond appropriately may result in an enformement action, such as an order for suspension or revocation of a license. When appropriate, prior to issuing a bulletin, the NRC may seek comments on the matter from the industry ( Atomic Industrial Forum, Institute of Nuclear Power Operations, nuclear steam suppliers, vendors, etc.), a technique which has proven effective in bringing faster and better responses from licensees. Bulletins generally require one-time action and reportirg. They are not intended as substitutes for revised license conditions or new requirements.

Infomation Notices are rapid transmittals of infomation 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 infomation to their facility.

Date Bulletin Issued Subject 83-02 3/4/83 STRESS CORROSION CRACKING IN LARGE-DIAMETER STAINLESS STEEL RECIRCULATION SYSTEM PIPING AT BWR PLANTS This bulletin informed all licensees and construction pemit holders of boiling water reactor (BWR) plants of recent generic pipe cracking problems. Extensive intergranular stress corrosion cracking in the main recirculation system piping was discovered during refueling outages at several plants. The effective-ness of ultrasonic testing (UT) methods used for inservice inspection of stainless steel BWR pipe welds, particularly in large-diameter piping, has prompted additional requirements. Licensees of 14 operating BWRs were required to demonstrate the effectiveness of the detection capability of UT methodology to examine welds in recirculation system piping before restaning scheduled operation.

All other licensees and construction pemit holders were infomed without additional action required.

___ Date Bulletin Issued Subject 83-03 3/10/83 CHECK VALVE FAILURES IN RAW WATER COOLING SYSTEMS OF DIESEL GENERATORS Information on check valve failurt was sent to (1) notify all licensees and construction pemit holders about numerous incidents of failed check valves in systems important to safety, (2) infom licensees of a significant generic matter for which additional NRC action is anticipated, and (3) require appropriate surveillance and testing of check valves in raw water cooling systems for diesel generators.

l Based on valve failures at the Dresdcn and Quad-Cities units, the following actions were required for holders of operating licenses:

(1) review and modify, if necessary, plant Ptap and Valve In-Service Test Program required by Section XI of the ASME Boiler and Pressure Vessel Code to include check valves in the flow path of cooling water for the diesel generators from the intake to the discharge; (2) verify valve integrity by identified procedures effective April 1, 1983 by the end of the next refueling outage; (3) report to NRC within 90 days of bulletin issuance identifying specific valves and describing verifica-tion procedures with proposed schedules; and (4) report to NRC within 90 days on completion of initial valve integrity verification procedures.

83-04 3/11/83 FAILURE OF THE UNDERV0LTAGE TRIP FUNCTION OF REACTOR TRIP BREAKERS Nuclear power facilities with operating pressurized water reactors were required to assure the proper operation of reactor trip breakers (except for facilities using Westinghouse DB type of breakers which were covered in Bulletin 83-01; see Power Reactor Events, Vol. 5, No. 1, pp. 28-29) as a result of failure of reactor trip breakers to open during testing at the San Onofre Nuclear-Generating Station. Other licensees and construction pemit holders were also infomed of this problem. The affected licensees were to perfom surveillance testing of the undervoltage trip function independent of the shunt trip function within five days of receipt of the bulletin, unless testing had already been performed within the last ten days, or before restning plant operation if in a shutdown mode. Further action required applicable licensees to report testing results and maintenance performed in response to this bulletin within ten days of receipt.

_ _ _ _ _ _ _ _ _ _ _ _ Infomation Date Notice Issued Subject 83-07 3/7/83 NONCONFORMITIES WITH MATERIALS SUPPLIED BY TUBE LINE CORPORATION All nuclear power fuel facilities and reactor facilities holding an operating license or con-struction permit were notified that pipe fittings and base flange forgings had been shipped from an anapproved nuclear source, had not been properly heat treated, and failed to meet ASME Code requirements for nondestructive examinations. Babcock and Wilcox (B&W) identified two heats of materials with strengths lower than that reported in the materials certification. The first procurement of the material in question was during 1980, and the first shipments for nuclear application were made during 1982.

83-08 3/9/83 COMPONENT FAILURES CAUSED CY ELEVATED DC CONTROL VOLTAGE All holders of a nuclear power reactor operating license or construction permit were notified of the generic implications inherent in three separate events that occurred at Trojan, Ft. Calhoun, and Zi on. These events indicated that de safety-related control components and indicating circuit components operating for a sustained period of time at elevated voltages or at above the rated design voltage degraded prematurely, causing short circuits and control problems.

83-09 3/9/83 SAFETY AND SECURITY OF IRRADIATORS All irradiator licensees were reminded that mechan-ical failures or htaan errors can result in serious, even fatal, overexposure to the radiation source.

Licensees were advised to emphasize to their employees that all available infomation relative to the position of the source should be checked before an employee enters the exposure room.

In addition, licensees should review their security programs to ensure their facilities are secured against unauthor-ized access at all times.

83-10 4/11/83 CLARIFICATION OF SEVERAL ASPECTS RELATING TO USE OF NRC-CERTIFIED TRANSPORT PACKAGES Specific matters relating to temporary shielding, preparation and assembly, quality assurance, and contamination surveys of NRC-Certified transport packages were clarified for all NRC-licensed reactor facilities and registered users of those packages.

___ _ _ _ _ _ _ _ _ _ _ _ Information Date Notice Issued Subj ect 83-11 4/14/83 POSSIBLE SEISMIC VULNERABILITY OF OLD LEAD STORAGE BATTERIES All nuclear power plant facilities holding an operating license or construction pennit were informed that pattern is developing of spontaneous failure of ola (therefore brittle) lead-acid storage batteries sug-gesting that a seismic event could cause a common-mode failure of plant DC systems. Also, material can slough off the plates, shorting out the battery or reducing its capacity. An event at Haddam Neck in 9/82 was discussed, and licensee event reports were listed for six similar events occurring at other plants.

Battery manufacturers mentioned were Gould and CSD, Division of ELTRA.

83-12 4/18/83 INCORRECT BORON STANDARDS All nuclear power reactor facilities holding an operating license or construction pennit were notified that the concentration deficiency of the boron standard of Lot No. 133107 supplied by the J. T. Baker Chemical Co. could result in an overestimation of the boron concentration in the primary coolant system. Such overestimation could reduce the capability to shut down the reactor and maintain the plant in a safe condition. The problem was discovered by the licensee for Palisades. Distribution of Lot No. 133107 was from four supply houses listed in the notice.

83-13 4/21/83 DESIGN MISAPPLICATION OF BERGEN-PATERSON STANDARD STRUT RESTRAINT CLAMP All nuclear power reactor facilities holding an operating license or construction permit were notified of a design misapplication of a standard strut restraint clamp (EA-3), provided by Bergen-Paterson Pipesupport Corp.

Sketches show the clamp, the acceptable application, and the misapplication.

Clanps improperly applied (with no shear lugs present) have substantially lower capabilities and cannot sustain the catalog load ratings without excessive deflection. Plants with the EA-3 clamps are listed.

83-14 4/21/83 DEWATERED SPENT ION EXCHANGE RESIN SUSCEPTABILITY TO EX0 THERMIC CHEMICAL REACTION All nuclear power reactor facilities holding an operating license or construction permit were provided early notification about the apparent

l h Information Date Notice Issued Subject susceptability of certain dewatered spent ion exchange resin to undergo exothermic and possibly explosive chemical reaction. The concern was based on an exothermic reaction that had occurred at Arkansas Nuclear One Unit 2 in January 1983. The organic material in question, stored for shipnent to a burial site, had most probably been exposed to chemicals and contaminants not normally asso-ciated with power operation. Nitrites and nitrates present in the affected resin apparently contributed to the exothermic reaction.

83-15 4/23 FALSIFIED PRE-EMPLOYMENT SCREENIL3 RECORDS All nuclear power reactor facilities holding an operating license or construction permit were alerted to the possibility that contractors might submit falsified security records to meet their commitments to the NRC. Audits, random revalida-tion, and random personal observation are some actions that have been taken to verify the quality of the process for screening security personnel.

83-16 4/30/83 CONTAMINATION OF THE AUBURN STEEL CORPANY PROPERTY WITH COBALT-60 All material licensees were reminded that, as a means of protecting the health and safety of the public, they must control licensed material and secure it against unauthorized removal. Licensees were advised to review provisions of the regulations and their license that deal with control of licensed materials and reporting lost or stolen material.

Those licensed to possess a Co-60 source were advised to review all receipt and transfer records, accounting for all sources. Pursuant to 10 CFR 20.402, any source not accounted for should immediately be reported to NRC.

83'-17 3/31/83 ELECTRICAL CONTROL LOGIC PROBLEM RESULTING IN INOPERABLE AUTO-START OF EMERGENCY DIESEL GENERATOR UNITS All nuclear power reactor facilities holding an operating license or construction permit were notified of a potential problem in the control logic circuitry which could adversely affect the auto-start feature provided for diesel generators

__ Infomation Date Notice Issued Subject at nuclear facilities. The control logic problem has strong generic implications.

Since the local /

remote control switch is located in the control room and is placed in the " remote" position during nomal plant operation, the licensee must be aware that direct immediate operator action is required to re-start the diesel generator ur.dar the set of conditions described in this infomation notice.

83-18 4/1/83 FAILURES OF THE UNDERVO.TAGE TRIP FUNCTION OF REACTOR TRIP SYSTEM BREAXERS All nuclear power reactor facilities holding an operating license or construction pemit were infomed about the need for regular, careful maintenance of reactor trip system (RTS) breakers to prevent failures of RTS circuit breakers with undervoltage (UV) trip attachments. New maintenance procedures are being prepared to reduce the likelihood of RTS breaker failure.

83-19 4/5/83 GENERAL ELECTRIC TYPE HFA RELAY CONTACT GAP AND 4

WIPE SETTING ADJUSTMENTS All holders of a nuclear power reactor operating license or construction pemit were notified that improper relay contact wipe and gap adjustments can affect the relay performance <>f the GE Type HFA relay during seismic events. Booklets are available to instruct the licensees on proper adjustment of this y

rel at.

83-20 4/13/83 ITT GRINNELL FIGU2E 306/307 MECHANICAL SNUBBER ATTACHMENT INTERFERENCE Ail nuclear power reactor facilities holding an operating license or construction pemit were informed about a potentially significant attactinent inter-ference problem in some ITT Grinnell Figure 306/307 mechanical snubber assemblies produced and shipoed betw en October 1978 and April 1980. Slight grinding (to ITT Grinnell specifications) in the area of inter-ference has alleviated the problem.

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Infomation Date Notice Issued Subject 83-21 4/15/83 DEFECTIVE EMERGENCY-USF. RESPIRATOR All nuclear power reactor facilities holding an operating license or construction permit, research and test reactors, fuel cycle licensees, and Priority 1 material licensees were informed about disintegrated " scrubber pads" in some BioPak 60P respirators manufactured by Rexnord, Inc. Use of the defective respirators could significantly impact the life and health of the user.

83-22 4/22/83 BOILING WATER REACTOR SAFETY / RELIEF VALVE FAILURES All boiling water reactor facilities holding an operating license or construction pennit were notified about failures of the pilot inlet tube attachments and problems of setpoint drift in Target Rock two-stage safety / relief valves, model number 7567F.

83-23 4/25/83 INOPERABLE CONTAINMENT ATMOSPHERE SENSING SYSTEMS All nuclear power reactor facilities holding an operating license or construction permit were reminded that management must give sufficient attention to maintaining containment integrity and to the operability of equipment related to the containment function.

Inadequate surveillance pro-cedures or inadequate implementation of the procedures, or both, were responsible in the examples c1ted for the existence of anomalous conditions that went undetected for too long.

83-24 4/28/83 LOOSE PARTS IN THE SECONDARY SIDE OF STEAM GENERATORS AT PRESSURIZED WATER REACTORS All pressurized water reactors holding an operating license or a construction permit were notified that loose parts on the secondary side of steam generators have been implicated in tube rupture events and in tube damage. The consequeaces of loose parts present some risk to public health and safety as well as an economic impact.

83-25 4/28/83 STANDBY GAS TREATMENT SYSTEM HEATER HIGH TEMPERATURE TRIP SETPOINT ADJUSTMENT All nuclear power reactor facilities holding an operating license or construction pennit were notified about a potentially significant problem pertaining to

_ Information Date Notice Issued Subject the standby gas treatment (SBGT) system. Latent heat from the heater and too low a trip setpoint caused a high-temperature trip of the heater, rendering the SBGT system inoperable. Repositioning the sensor and installing an automatic reset device are among the solutions offered.

l 2.3 Engineering Evaluations and Case Studies Issued in March - April 1983 The Office for Analysis and Evaluation of Operational Data ( AE0D) has as a primary responsibility the task of reviewing the operational experience reported by NRC nuclear power plant licensees.

As part of fulfilling this task, it selects events of apparent interest to safety for further review as either an engineering evaluation or a case study. An engineering evaluation is usually an immediate, general consideration to assess whether or not a more detailed, protracted case study is needed. The results are generally short reports, and the effort involved usually is a few staffweeks of investigative time.

Case studies are in-depth investigations of apparently significant events or situations. They involve several staffmonths of engineering effort, and result in a fomal report identifying the specific safety problems (actual or potential) illustrated by the event and recommending actions to improve safety and prevent recurrence of the event. Before issuance, this report is sent for peer review and comment to at least the applicable utility and appropriate NRC offices.

These AEOD reports are made available for information purposes and do not impose any requirements on licensees.

The findings and recommendations contained in these reports are provided in support of other ongoing NRC activities concerning the operational event (s) discussed, and do not represent the position or requirements of the responsible NRC program office.

Case Date Study Issued Subject C301 4/83 FAILURE OF CLASS 1E SAFETY-RELATED SWITCHGEAR CIRCUIT BREAKERS TO CLOSE ON DEMAND Data were reviewed concerning 108 failures of Class 1E safety-related switchgear circuit breakers to close on demand. These events occurred at 47 plants during approximately five and one-half years.

In general, failures were attributed to a problem within the attendant closing circuit such as blown control fuses, intermittent electrical connections, dirty or corroded contacts, and malfunctions in the spring charging motor or associated limit switch contacts. Electrical circuit problems for a given circuit breaker were found to be repetitive, and few problems were directly attributed to a specific mechanical problem. Nearly 25% of the events involved a diesel generator output breaker.

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-. - - - - _ - _ Engineering Date Evaluation Issued Subject E304 3/11/83 INVESTIGATION OF BACKFLOW PROTECTION IN COMMON EQUIPMENT AND FLOOR DRAIN SYSTEMS TO PREVENT FLOODING 0F VITAL EQUIPMENT IN SAFETY-RELATED COMPARTMENTS On 11/5/81, Bechtel Corporation notified the licensee for Calvert Cliffs 1 and 2 that the watertight integrity of the service water pep rooms at both units could be impaired because check valves had not been installed in the floor drain system which drains by gravity to the turbine condenser pit in the turbine building. Without these check valves, the operability of the service water ptsnps could not be assured in the event of a circulating water conduit break in the turbine building (a nonsafety area). As a temporary measure, the licensee sealed some of the drains with inflatable plugs and modified the remaining drain lines by installing check valves. The AE0D staff investigated NRC requirements on the protection of safety-related equipment from possible backflow in the floor drainage system.

E305 4/13/83 INOPERABLE MOTOR-0PERATED YALVE ASSEMBLIES DUE TO PREMATURE DEGRADATION OF MOTORS AND/0R IMPROPER LIMIT SWITCH / TORQUE SWITCH ADJUSTMENT This report evaluates eight licensee event reports involving motor-operated valves:

four at Vennont Yankee, two at San Onofre 2, one at James A. Fitzpatrick, and one at Oconee 1.

These events illustrate several modes of valve assembly failure to operate. The cited contributors to inoperability are (1) component (or system) cooldown, (2) inadequate adjustment (and inadequate switch, (3) procedures) of a limit switch and torque inadequate motor thermal overload pro-tective devices, and (4) vibration-related phenomena that apparently led to loosening of a liinit switch with subsequent excessive motor torque or motor burnout.

Some modes of failure, such as vibration-related phe-nomena, do not appear to have been anticipated or accomodated in design, qualification, or inservice test programs.

E306 4/14/83 C00LDOWN DURING LOSS OF CONTROL ROOM TEST On 9/14/81, while perfoming a loss-of-control-room test during startup testing at McGuire 1, plant operators had difficulty controlling the auxiliary

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Engineering Date Evaluation Issued Subject feed system, resulting in an abnomally fast cooldown rate and a safety injection. The greatest contributing factor to the transient was the operators' inability to control auxiliary feedwater addition. Al so, inade-quate procedures failed to indicate that all operators should be present at the auxiliary shutdown panels before the reactor was tripped. The lack of reset switches for the auxiliary feed control valves outside the control room indicated a design deficiency.

The licensee installed duplicate reset switches in the auxiliary shutdown panels. Procedures were modified to require depressing the auxiliary flow reset switches before leaving the control room, and to require the operators of the reactor trip-to-breaker panel and the auxiliary control panel to establish communication before tripping the reactor. A second test was perfomed on 9/17/81 using the revised pro-cedures. The problems of the earlier test did not appear.

E307 4/18/83 DEGRADATION OF SAFETY-RELATED BATTERIES DUE TO CRACKING OF BATTERY CELL CASES AND/0R OTHER POSSIBLE AGING-RELATED MECHANISMS Battery-related events at three plants were reviewed:

Indian Point 2, Haddam Neck, and North Anna 1.

At North Anna 1, a battery failed its 18-math discharge surveillance test. At the other two plants there were leaking cells, low electrolyte levels, cracked cell casings, and expansion of cell plates. The question remains as to whether the batteries would have perfomed their safety function on demand in the "as found" condition.

Such physical degradation of safety-related battery cells at or near their expected end of battery cell life could result in a common mode failure of a group of battery cells.

Actual replacement criteria may not give adequate consideration to related aging mechanisms which could affect their functional perfomance on demand.

Inadequate margin between applied load testing curves used for service testing and battery design rating may also exist at other nuclear facilities.

This condition results in accelerated aging with an attendant decrease in battery capacity, thus requiring more frequent battery replacements.

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_ _ _ _ _ _ _ _ Engineering Date Evaluation Issued Subject E308 4/19/83 CRACKS AND LEAKS IN SMALL DIAMETER PIPING Three separate leaks at weld penetrations involving small piping in the makeup system at Crystal River 3 were described. A literature search revealed that between 1/82 and 11/82, 34 similar events (including the three at Crystal River) that involved 22 units were reported. Of these 22 units, Dresden 3 had six events, Browns Ferry had four, and the remaining 20 units had one or two reports with a single leak event for each report. Most of the leaks were minor; however, Big Rock Point, Oyster Creek, and Browns Ferry leaked 1,000, 9,700, and 17,200 gallons, respectively.

Small-pipe leaks seem to be related to unanticipated vibration problems that require t 'e addition of supports or rerouting of original piping. Leaks also have occurred when supports added as a corrective action (1) were later removed to perform maintenance while a system was still in operation, or (2) were not reinstalled subsequent to maintenance because they were not shown on the design drawings. About 20%

of the leakage events involved threaded connections.

Some leaks were located in portions of lines that were unisolable. These events represent a potential generic problem in that operation-induced vibration loads were not part of the design specification for these piping systems.

E309 4/21/83 THE P0TENTIAL FOR WATER HAMMER DURING THE RESTART OF THE RHR PUMPS AT BWR NUCLEAR POWER PLANTS A concern was addressed relating to postulated scenarios of water hammer in the residual heat removal (RHR) system of a boiling water reactor (BWR) caused by void formation due to interruption of RHR punp operation while in certain containment spray or cooling modes.

Browns Ferry 1 is considered as a prototype for system arrangement, response, and operational analyses. The results of the evaluation suggest that the possibility of void fomation in the RHR system and subsequent water hammer should be considered by BWR plant operations personnel for all BWRs when developing RHR system operating

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Engineering Date Eval ua_ tion Issued Subj ect E310 4/25/83 LOSS OF SHUTDOWN COOLING AND SUBSEQUENT BORON DILUTION AT SAN-0N0FRE-2 During preoperational testing on 3/14/82, the shutdown cooling (SDC) system at the San Onofre Unit 2 was inoperable for 90 minutes when the low pressure safety injection (LPSI) pumps became nitrogen bound. The event resulted from an improper valve alignment during required nitrogen backflushing of a purificatian filter in the chemical and volume control system (CVCS). As a result of either a system malfunction or operator error, gaseous nitrogen passed through the purifica-tion line into the suction of the LPSI punps which were being used for shutdown cooling SDC flow was restored within 90 minutes. Because back-flushing of the purification filter is a typical operation for all PWRs, the loss of SDC identified a potential safety concern if similar circumstances should occur for even relatively brief periods of time with an irradiated core.

E311 4/25/83 LOSS OF SALT WATER FLOW TO THE SERVICE WATER HEAT EXCHANGERS FOR 23 MINUTES AT CALVERT CLIFFS UNIT 2 On 6/20/82, Calvert Cliffs Unit 2 experienced a loss of both redundant trains of the safety-related service water system when a single nonsafety-related butterfly valve in the salt water system failed and blocked all salt water coolant fl ow. Since this event was not considered to be credible in the safety reviews, the operating procedures did not include corrective measures for the possibility of system flow bicckage. The licensee, however, diagnosed the problem and restored flow to one of the service water heat exchangers before any temperature alarms actuated.

The actual consequences of the event to plant equipment were minimal.

The AE0D staff concludad that the system design could be improved from a safety standpoint by evaluating the source of potential single failure which can cause loss of the system function, i.e.,

the nonsafety-related valves in the common discharge headers from the service water and component cooling water heat exchangers. The licensee is considering this option. A review of the FSARs of other operating pressurized water reactors indicated that Calvert Cliffs is not unique in having a single valve in the common discharge header of redundant safety-related heat exchangers.

_ _ _ - _ _ - _ - _ _ _ - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ - _ - _ - - - - - 2.4 Generic Letters Issued in March-April 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 infomation to, and obtain infornation from, reactor licensees, applicants, and/or equipment suppliers regarding matters of safety, safeguards, or environ-mental significance. During March and April 1983, six 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 nunber of important issues, the resolutions of which have contributed to improved quality of design and operation.

Generic Date letter Issued Subject 83-13 3/2/83 CLARIFICATION OF SURVEILLANCE REQUIREMENTS FOR HEPA FILTERS AND CHARC0AL ADSORBER UNITS IN STANDARD TECHNICAL SPECIFICATIONS ON ESF CLEANUP SYSTEMS All applicants for operating licenses and holders of construction pemits for power reactors were notified that the wording in the surveillance requirements for testing the high-efficiency particulate air (HEPA) filters and charcoal adsorber units had been modified.

This was done to clarify the required relationship between the guidance provided in Regulatory Guide 1.52, Rev. 2 and ANSI N510-1975; the testing re-quirements of the HEPA filters and charcoal adsorber units; and the assunptions used by the NRC staff in its safety evaluations for the engineered safety feature (ESF) atmospheric systems.

The revised require-ments m.re er. closed.

83-14 3/7/83 DEFINITION OF "XEY MAINTENANCE PERSONNEL" (CLARIFICATION OF GENERIC LETTER 82-12)

All licensees of operating reactors, applicants for operating licenses, and holders of construction pemits were advised that key maintenance personnel are defined as "those personnel who are responsible for the correct performance of maintenance, repair, modification or calibration of safety-related structures, systems or components, and who are l

personnel perfoming or immediately supervising the performance of such activities."

- - - - - _ - _ - _ _ _ _ _ _ _ - _ _ _ _ Generic Date Letter Issued Subject 83-15 3/23/83 IMPLEMENTATION OF REGULATORY GUIDE 1.150, " ULTRA-SONIC TESTING OF REACTOR VESSEL WELDS DURING PRESERVICE AND INSERVICE EXAMINATIONS," REVISION 1 All licensees of operating power reactors and applicants for operating licenses were advised that Revision 1 of Regulatory Guide (RG) 1.150 incorporates an alternative ultrasonic testing method recommended by the Ad Hoc Committee of the Electric Utility Industry. If appropriate, licensees should modify their technical specifica-tions to make them consistent with Rev.1 of RG 1.150.

83-16 3/24/83 TRANSMITTAL OF NUREG-0977 RELATIVE TO THE ATWS EVENTS AT SALEM GENERATING STATION, UNIT NO. 1 All light water plant licensees and applicants were notified that a task force from NRC Region I has documented its investigation of the 2/22/83 and 2/25/83 anticipated' transient without scram

( ATWS) events at Salem 1 in NUREG-0977. Recipients of this letter should ensure that the report is made available to plant staff as part of the plant training program.

83-17 4/8/83 INTEGRITY OF THE REQUALIFICATION EXAMINATIONS FOR RENEWAL OF REACTOR OPERATOR AND SENIOR REACTOR OPERATOR LICENSES All power and non-power reactor licensees, applicants for an operating license, and holders of a construction permit were advised to ensure that the validity of the certifications required by 10 CFR Part 55 and the integrity and honesty of the licensed operator requalification program are adequately addressed in facility procedures. Procedures should guarantee that cheating will either be prevented or, if it occurs, will be detected.

83-18 4/19/83 NRC STAFF REVIEW 0F THE BWR OWNERS GROUP (BWROG)

CONTROL ROOM SURVEY PROGRAM All boiling water reactor licensees of operating reactors, applicants for an operating license, and holders of construction pennits were provided with the results of the NRC staff's evaluation 3

of the BWR Owners Group (BWROG) Control Room Survey Program. Recipients of this generic letter were expected to submit a plant-specific i

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- r Generic Date Letter Issued Subject program plan to the NRC referencing the BWROG Generic Program Plan, and to complete the BWROG control room survey Checklist Supplement.

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_ 2.5 Operating Reactor Event Memoranda Issued in March-April 1983 The Director, Division of Licensing, Office of Nuclear Reactor P. gulation (NRR),

disseminates information to the directors of the other divisions a.a program offices within NRR via the operating reactor event memorandum (OREM) system.

The OREM documents a statement of the problem, background infonnation, the safety significance, and short and long tenn 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.

No OREMs were issued during March-April 1983.

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- 2.6 Regulatory and Technical Reports Issued in March-April 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 ab* tracts 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, PHI'.-016, Washington, DC 20555 or on (301) 492-9530.

Report Title NUREG-0020 LICENSED OPERATING REACTORS STATUS

SUMMARY

REPORT (DATA AS Vol. 6, No. 12 0F NOVEMBER 30, 1982)

April 1983 This report provides data on the operation of nuclear units as timely and accurately as possible.

The infomation is collected by the Office of Management and Program Analysis, 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-0390 TOPICAL REPORT REVIEW STATUS (DATA AS OF JANUARY 20, 1983)

Vol. 6, No. 2 March 1983 The primary purpose of this report is to provide periodic progress reports of ongoing topical report reviews, to identify those topical reports for which NRC staff review has been completed and, to the extent practicable, to prcvide NRC management with sufficient information regarding the conduct of the topical report program to pemit taking whatever actions are deemed necessary or appropriate.

NUREG-0540 TITLE LIST OF DOCUMENTS MADE PUBLICALLY AVAILABLE (JANUARY 1-31, Vol. 4, No. 1 1983)

April 1983 This document is a monthly publication containing descriptions of infomation received and generated by the NRC. This infor-mation 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.

________ __ t Title Report NUREG-0728 NRC INCIDENT RESPONSE PLAN Rev. 1 April 1983 The NRC regulates civilian nuclear activities to protect the public health and safety and to preserve environmental quality. An Incident Response Plan had been developed and has now been revised to reflect current Commission policy. NUREG-0728, Rev.1 assigns responsibilities for responding to any potentially threatening incident involvir.g NRC-licensed activities and for assuring that the NRC will fulfill its statutory mission.

NUREG-0897 CONTAINMENT EMERGENCY SUMP PERFORMANCE April 1983 This report summarizes key technical findings related to the Unresolved Safety Issue (USI) A-43, " Containment Emer-gency Sump Perfomance." The technical issues can be sub-divided into sump hydraulic performance, loss-of-coolant accident (LOCA) generated debris effects, and recirculation pep perfomance under post-LOCA conditions. The technical findings presented in this report provide a means to address these smp design and perfomance aspects, and provide the technical basis for the proposed changes to NRC s Standard Review Plan, Section 6.2.2, and Regulatory Guide 1.82.

NUREG-0940 ENFORCEMENT ACTIONS: SIGNIFICANT ACTIONS RESOLVED Vol. 2, No. 1 April 1983 This compilation samarizes significant enforcement actions that have been resolved during one quarterly period (January-March 1983) and includes copies of letters, notices, licensee responses, and orders sent by the NRC to the licensees with respect to the enforcement actions.

It is anticipated that the infomation in this publication will be widely dis-seminated to managers and employees engaged in activities licensed by the NRC, in the interest of promoting public health and safety as well as common defense and security.

NUREG-0977 NRC FACT-FINDING TASK FORCE REPORT ON THE ATWS EVENTS AT March 1983 SALEM NUCLEAR GENERATING STATION, UNIT 1, ON FEBRUARY 22 AND 25, 1983 An NRC Region I Task Force was established on March 1, 1983 to conduct fact finding and data collection with regard to the circumstances which led to the anticipated transient without scram ( ATWS) events at the Public Service Electric and Gas Company's Salem Generating Station, Unit 1, on February 22 and 25,1983. The charter of the Task Force was to determine the factual infomation pertinent to management and administrative controls which should have ensured proper operation of the reactor trip circuit breakers in the solid state protection system. This report documents the findings of the Task Force along with its conclusions and recommenda-tions.

Report Title NUREG-1000 GENERIC IMPLICATIONS OF ATWS EVENTS AT THE SALEM NUCLEAR Vol. 1 POWER PLANT April 1983 This report is the first of two volumes.

It documents the work of an interoffice, interdisciplinary NRC Task Force established to detennine the generic implications of the two anticipated transients without scram (ATWS) at the Salem Nuclear Power Plant, Unit 1, on February 22 and 25,1983.

A second report will document the NRC actions to be taken based on the work of the Task Force. The Task Force was established to address three questions:

(1) Is there a need for prompt action for similar equipment in other facilities? (2) Are NRC and its licensees learning the safety-management lessons? and, (3) How should the priority and content of the ATWS rule be adjusted? A number of short-tenn actions were taken through bulletins and an infonnation notice.

Intermediate-tenn actions to address the generic issues will be addressed in the separate report and implemented through appropriate regulatory mechanisms.

Ultimately, the regulatory and programmatic changes will be incorporated into the Regulations, Standard Review P1an, manual chapters, and other documents as necessary to assure continued attention to the lessons learned from the Salem Unit 1 ATWS events.

NUREG/CP-0035 THE MAN-MACHINE INTERFACE AND HUMAN RELIABILITY: AN April 1983 ASSESSMENT AND PROJECTION In December of 1979, with the incident of TMI-2 still fresh in the mind of nuclear industry, a meeting was held in Myrtle Beach, South Carolina, to address the role of the human in the safety of nuclear power plant operations.

This interdisciplinary meeting was sometimes marked by rather heated discussions as representatives from each technical area groped their way through the meeting sessions.

Despite this, it may well be stated that the meeting marked the beginning of a meaningful dialogue between the disciplines in order to allow the questions to be properly fonnulated.

If Myrtle Beach I can be described as a pioneering effort, then Myrtle Beach II was a settlement effort. The scope of the second meeting, recorded herein, encompassed the signif-icant short-term (notably control room reviews, and safety parameter display systems) and longer tenn issues (such as the integration of hunan factors in the entire design process, and the use of automated control features). Specific and

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significant agreements and recommendations resulted from the week's work in these areas. In the area of human perfonnance, which was more in line with the scope of the original meeting, a need was shown for the development of a taxonomy or model

_ - _ _ _ _ _ _ Report Title to structure future data gathering, and the need for models and data to address the issue of cognitive behavior.

NUREG/CR-2000 LICENSEE EVENT REPORT (LER) COMPILATION (VOL. 2, NO. 2 FOR Vol. 2, No. 2 FEBRUARY 1983; VOL. 2, NO. 3 FOR HARCH 1983)

March 1983; Vol. 2, No. 3 This monthly report contains Licensee Event Report (LER)

April 1983 operational information that was processed into the LER data file of the Nuclear Safety Information Center (NSIC) during the one month period identified on the cover of this document. The LERs, from which this information is derived, are submitted to the NRC by nuclear power plant licensees in accordance with Federal regulations.

Procedures for LER reporting are described in detail in NRC Regulatory Guide 1.16 and NUREG-0161, " Instructions for Preparation of Data Entry Sheets for Licensee Event Reports."

The LER sunmaries in this report are arranged alphabetically by facility nane and then chronologically by event date for each facility. Component, system, and keywords indexes follow the sunmarie:. The components and systems are those identified by the utility when an LER form is initiated; the keywords are assigned by the NSIC staff when the sunmaries are prepared for computer entry.

NUREG/CR-2005 CHECKLIST FOR EVALUATING EMERGENCY OPERATING PROCEDURES Rev. 1 USED IN NUCLEAR POWER PLANTS April 1983 This document describes a checklist to be used by NRC's Office of Inspection and Enforcement inspectors during their evaluation of emergency operating procedures. The objective of the checklist is to aid inspectors in identifying character-istics of the procedures that can lead to operator performance deviations. Explanations of the procedures evaluation criteria comprising the checklist are provided. Methods of performing the evaluations are described, and suggestions for applying the checklist to increase the effectiveness of the inspection process are made. A companion document, " Development of a Checklist for Evaluating Emergency Procedures Used in Nuclear Power Plants," NUREG/CR-1970, describes the methodology used to develop the checklist and presents the study findings on which the procedures evaluation criteria are based.

Revision 1 of the checklist, presented herein, is the result of a one-year field test by NRC inspectors in all five NRC regions.

It incorporates improvements that were suggested by inspectors based on their experience with the checklist in performing evaluations of licensee procedures.

Report Title NUREG/CR-2524 EVALUATION OF PERSONNEL NEUTRON 00SIMETRY AT OPERATING March 1983 NUCLEAR POWER PLANTS The basic objective of this research program titled,

" Evaluation of Personnel Neutron Dosimetry at Operating Nuclear Power Plants," sponsored by the NRC was to evaluate neutron personnel monitoring devices and/or methods used at nuclear power plants. To accomplish this research.

measurements were made in areas where personnel are likely to have access during reactor operation. Measurements were made of spectral and flux distribution of neutrons.

Dose equivalent surveys were made of neutrons and gammas in operating areas with portable survey meters. Gamma to neutron ratios were obtained. A new computer code, BONABS, was developed to incorporate the response to portable neutron survey meters with an existing code, BONHER, used for neutron spectrtan unfolding techniques.

NUREG/CR-2531 INTRODUCTORY USERS MANUAL FOR THE U.S. NUCLEAR REGULATORY Rev. 1 COMMISSION REACTOR SAFETY RESEARCH DATA BANK March 1983 The NRC has established the NRC/ Division of Accident Evaluation (DAE) Data Bank Program to collect, store, and make available data from the many domestic and foreign water reactor safety research programs. The NRC/DAE Data Bank Program provides a central computer storage mechanism and access software for data that is to be used by code development and assessment groups in meeting the code and correlation needs of the nuclear industry. The administrative portion of the program provides data entry, docmentation, training, and advisory services to users and the NRC. The NRC/DAE Data Bank and the capabilities of the data access software are described in this document.

NUREG/CR-2728 INTERIM RELIABILITY EVALUATION PROGRAM PROCEDURES GUIDE March 1983 This document presents procedures for conducting analyses of a scope similar to those perfonned in Phase II of the Interim Reliability Evaluation Program (IREP).

It doctanents the current state of the art in perfonning the plant systems analysis portion of a probabilistic risk assessment.

Insights gained into managing such an analysis are discussed. Step-by-step procedures and methodological guidance constitute the major portion of the document. While not to be viewed as a " cookbook," the procedures set forth the principal steps in perfonning an IREP analysis. Guidance for resolving the problems encountered in previous analyses is offered.

Ntmerous examples and representative products from previous i

analyses clarify the discussion.

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__ Title Report NUREG/CR-2956 NEUTRON DOSIMETRY AT COMMERCIAL NUCLEAR PLANTS March 1983 As part of a larger program to evaluate personnel neutron dosimetry at commercial nuclear power plants, this study was designed to characterize neutron dosimeter responses inside the containment structure of canmercial nuclear pl ants.

In order to characterize those responses, dosimeters were irradiated inside containment at two pressurized water reactors (PWRs) and at pipe penetrations outside the biological shield at two boiling water reactors (BWRs) while the reactors were operating at full power. Additionally, the dosimeters were irradiated (1) using monoenergetic neutrons produced by an accelerator; and (2) using the filtered reactor irradiations. Simultaneous measurements were taken using a tissue equivalent proportional counter and portable remmeters, SN0OPY, RASCAL and PNR-4.

The results of the analyses of dosimeter responses indicate that (1) the dosimeters irradiated inside containment of PWRs respond as if the dosimeters were irradiated using monoenergetic neutrons below 100 kev; (2) that the use of bare neutron sources for dosimeter calibrations is inappropriate for the in-containment irradiations; (3) the TLD-albedo dosimeter is the only type of dosimeter available that demonstrated both adequate precision and sensitivity; (4) that the polycarbonate track etch dosimeter which uses radiators was sensitive enough, but demonstrated inadequate precision for the reactor irradiations; and that (5) CR-39 and poly-cabonate track etch dosimeters which do not use radiators were inadequate for use inside. containment of nuclear power pl ants.

NUREG/CR-3059 PARAMETRIC CALCULATIONS OF FATIGUE CRACK GROWTH IN PIPING March 1983 This study present calculations of the growth of piping flaws produced by fatigue. Flaw growth was predicted as a function of the initial flaw size, the level and nunber of stress cycles, the piping material, and environmental factors.

The results indicate that the present flaw acceptance standards of ASME Section XI provide a relatively consistent set of allow-able flaw sizes because the predicted life of flawed piping is relatively insensitive to pipe wall thickness, flaw aspect ratio, and piping material (ferritic versus austenitic). On the other hand, the results show that flaws acceptable under ASME Section XI can grow at unacceptable rates if the cyclic stresses are at the maximum level permitted by the design rules of ASME Section III. However, a review of the conser-vatisms inherent to the ASME code rules is presented to explain the low occurrence of piping fatigue failures in service.

It is concluded that decreases in the allowable flaw sizes are not justified.

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