Text

Effects of Ambient Temp on Electronic Components in Safety-Related Instrumentation & Control Sys
	ML20215E892
Person / Time
Issue date:	12/31/1986
From:	Chiramal M; NRC OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL DATA (AEOD)
To:	;
Shared Package
ML20215E878	List: ML20215E875; ML20215E892;
References
	TASK-AE, TASK-C604 AEOD-C604, NUDOCS 8612230157
	Download: ML20215E892 (99)
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i CASE STUDY REPORT

AE0D/C604 EFFECTS OF AMBIENT TEMPERATURE ON ELECTRONIC COMPONENTS IN SAFETY-RELATED INSTRUMENTATION AND CONTROL SYSTEMS December 1986 Prepared by: M. Chiramal p ;

Office for Analysis and Evaluation of Operational Data ~ f U.S. Nuclear Regulatory Comission '

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This report documents the results of a study completed by the Office for Analysis and Evaluation of Operational Data with regard to selected operating events. The findings, conclusion, and recommendations contained in this l report are provided in support of other ongoing NRC activities and do not

represent the position or requirements of the responsible program office or the Nuclear Regulatory Commission.

8612230157 861205 PDR ORG NEXD PDR l

. d TABLE OF CONTENTS Page EXECUTIVE

SUMMARY

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1.0 INTRODUCTION

........................ 5 2.0 OPERATING EXPERIENCE .................... 6 2.1 Loss of Control Area Ventilation System at the McGuire Station .................... 6 2.1.1 Event Discussion ................ 6 2.1.2 Analysis and Evaluation of the Event . . . . . . 6 2.2 Pcwer Supply Failure at Davis-Besse . . . . . . . . . . . 9 2.2.1 Event Discussion ................ 9 2.2.2 Analysis and Evaluation of Event ........ 10 2.3 Failure in ESFAS Cabinet at Palo Verde Unit 1 . . . . . 10 2.3.1 Event Discussion ................ 10

.2.3.2 Analysis and Evaluation of the Event . . . . . . 12 2.4 Operational Experience at Summer 1 . . . . . . . . . . . 13 2.4.1 Event Discussion ................ 13 2.4.2 Analysis and Evaluation of Operational Experience ................... 15 2.5 Other Events . . . . . . . . . . . . . . . . . . . . . . 15 3.0 GENERIC APPLICABILITY OF THE PROBLEM 0F HIGH AMBIENT TEMPERATURE ..................... 17 4.0 STATION BLACK 0UT AND HIGH AMBIENT TEMPERATURES . . . . . . . . 20 5.0 FINDINGS AND CONCLUSIONS . . . . . . . . . . . . . . . . . . . 22 5.1 McGuire Station Event . . . . . . . . . . . . . . . . . . 22 5.2 Davis-Besse Event . . . . . . . . . . . . . . . . . . . . 23 5.3 Palo Verde 1 Event ................... 24 l

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TABLE OF CONTENTS (continued)

Page 5.4 Generic Aspects of the Problem ............. 24 5.5 Station Blackout .................... 25 6.0 RECOMMENDATIONS ....................... 26

7.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . . . 28 APPENDIX - Events Involving Failures in Westinghouse Solid State Protection and Process Control Systems . . . . . . 29

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i EXECUTIVE SUP94ARY This report documents the review and evaluation of four events involving failures of solid state electronic components in safety-related instrumentation and control systems due to overheating. Failures of electronic components in safety-related instrumentation' systems lead to malfunctioning of control systems, plant transients, inoperability of instrumentation channels in protection systems, inadvertent actuations and/or failures of safety systems, i

and erroneous-indications and alarms in the control room. Unless plant operators are made aware of and trained in the potential consequences and behavior of plant instrumentation systems exposed to elevated ambient temperatures, such failures, especially common cause failures, can lead to unsafe operation of the plant.

This review was initiated by an event that occurred at the McGuire Station in June 1984, during which the plant's control area ventilation system was inoperable for 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Based on this event, and a resulting initial AE00 study, Information Notice 85-89, " Potential Loss of Solid-State Instrumentation Following Failure of Control Room Cooling," was issued in November 1985, to alert licensees of operating nuclear plants of the event and concerns. Three other events were subsequently identified in which the root cause of failure of the instrumentation system was determined to_be overheating of electronic components located in cabinets in a controlled environmental area. These events occurred at Davis-Besse, Palo Verde 1, and Summer 1. These and the McGuire event are summarized below.

On June 4, 1984, while both units at the McGuire Station were operating at 100 percent power, a total loss of the control area ventilation system occurred. Approximately 45 minutes later, the control room received numerousalarmsonUnit1indicatinghigh' ) in reactor coolant loop C. Alarms were also receivedtemperaturefor pressuriz (T@'levelonUnit1.

One hour into the event the ventilation system was declared inoperable, and I hour after that the control room operators started to reduce power on

, both units as required by plant technical specifications. During this j period other alarms were also received on both units. At approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months into the event, the instrument cabinet doors were opened and cool air l from the air-conditioned computer room was moved through the control room doorway to the cabinets by portable fans and ducts. About 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the initial ventilation system loss, the control room area ventilation system

was returned to service and declared operable. By this time the control

room operators had reduced power on both units to 97 percent.

! Both McGuire units have previously experienced failures of instrumentation and control systems due to overheating of electronic components in the Process Control System (PCS 7300) and the Solid State Protection System (SSPS) cabinets located in the control room area. Thus, the operators anticipated the problem with the instruments and correctly assumed that the alarms they were receiving were spurious. In spite of previous experiences, it took the operators approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to provide alternate cooling to the affected instrumentation cabinets. The plant was operated safely during

the event due to the efforts of the operators in providing alternate cooling to the affected instrumentation cabinets and in not being confused by the j.

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, t spurious alarms in the control room. If such actions had not been taken, the electronic component failures, would likely have led to transients and trips of both units, inadvertent actuations and/or inoperability of safety-related systems, and additional spurious alarms and erroneous indica-tions in the control room.

Following the event, and with the control area ventilation system operating normally, the licensee measured the temperatures inside the PCS cabinet with the highest heat load. The temperatures obtained were found to be much higher (42 F to 52*F above room ambient) than the anticipated 20"F above ambient. After the incident, maintenance personnel rebalanced the air flow in the control area ventilation system to provide additional cooling to the instrumentation cabinets. The licensee has implemented other modifica-tions, such as modifying the heat sinks on the PCS cards and improving the operation of the chillers, to improve the reliability of the electronic components in the instrument cabinets located in the control area. It should be noted that the McGuire Station has been experiencing card failures and associated spurious instrument indicaticns since 1981, and yet only in 1984-1985 was the root cause of the failures fully identified and adequate corrective actions implemented.

On September 21, 1982, while Davis-Besse I was operating at full power, the control room operators received a full trip alarm on Channel 3 of the Steam and Feedwater Rupture Control System (SFRCS). Investigation of the trip identified the cause to be the actuation of the overvoltage trip device of the 48-volt dc/dc power supply for the field contacts input. The power supply was reset, and the input and output voltages were checked and found to be correct. However,15 minutes later the power supply failed again.

It was reset a second time, but tripped a few minutes later. After the third failure a new power supply was installed. The failed power supply was sent to the manufacturer for analysis.

In 1984, from the analysis results from the manufacturer, it was determined that inadequate design of the SFRCS cabinets had resulted in overheating of the power supplies in these cabinets, and this has been identified as a primary contributing factor for the problem of the power supplies. By a field change review, the licensee has installed fans in the SFRCS cabinets to ensure that adequate cooling is provided. The licensee is planning to measure the temperatures in the cabinets to verify adequacy of cooling.

On December 16, 1985, Palo Verde Unit I was operating at 52 percent reactor power when an electronic failure in a train A engineered safety features (ESF) cabinet resulted in the spurious actuation of several ESF signals. A spurious auto-start of the train A emergency diesel generator, operation of the ESF load sequencer, and a train A load shed signal also occurred as a result of the failure. The load shed signal did not clear, and did not allow the automatic or manual loading of electrical loads onto the safety-related bus and diesel generator.

The cause of the event has been traced to the failure of a fan in the ESF cabinet which allowed the ESF load sequencer module to overheat and malfunc-tion. As an immediate corrective action, the ESF sequencer module was replaced, the ESF cabinet door was temporarily removed, and the fan was repaired. A control room alarm which annunciates on high cabinet exit air temperature is planned to be installed.

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At Summer 1 on December 16, 1982, with the plant in Mode 1, the main control board indicator for reactor building pressure failed low. The plant has experienced similar events in the past. The licensee concluded that the cause of the events was the failure of a loop power supply circuit board when exposed to the ambient conditions present in the cabinet. As a temporary measure, the licensee installed additional room air-conditioners to alleviate the problem. However, on January 8, 1983, the main control room indication failed low for feedwater flow on one steam generator. The event was also attributed to high instrument cabinet temperature. Testing by the licensee indicated air stagnation in the relay room.

In April 1983, the licensee modified the relay room ventilation for more effective cooling of the instrument cabinets. Temperature measurements were taken of the ambient temperature at various locations in the relay room to assure adequacy of the modification. Since then, the licensee has not experienced failures of electronic components in the cabinets in the relay room attributable to elevated temperature.

The r"/iew of these events found that, in general, identifying elevated room ambient temperatures or instrument cabinet internal temperatures as the root cause for the failure of electronic components has not been immediate or easy.

Licensees, over an extended period, experienced several failures and many corrective actions before finally identifying overheating of components as the underlying reason for many of the failures they had experienced. Based on the type of component failures experienced at the McGuire Station and Summer 1, a limited search of the Sequence Coding and Search System (SCSS) for events involving similar failure modes at operati'ig Westinghouse PWR units was done.

Many such events were identified, but, in general, high room ambient temperature or instrument cabinet internal temperature was not identified as a possible root cause of the failures.

As a result of the events summarized above, technical specifications regarding area ventilation cooling systems and instrumentation systems also were reviewed, and were found to be inadequate with respect the temperature rating of electronic components in the instrument cabinets.

In addition, a review of the staft's proposed resolution of USI A-44 regarding design adequacy and capability of instrumentation and control system equipment needed to function in environmental conditions associated with a station blackout found that additional actions were warrented. Specifice.lly, plant specific evaluations are needed with regard to the actual tempera;ture and condition of heat sensitive components inside instrument cabinets.

Overheating of electronic components in safety-related instrumentation and control systems raises two concerns: (1) decreased reliability of electronic equipment due to increased failure rate of printed circuit cards and other heat sensitive electronic components, and (2) the potential for comon cause failure of redundant safety-related instrumentation channels due to extended loss of normal cooling air flow to the cabinets in which the instruments are located.

These concerns are generic to all operating nuclear units that utilize solid state electronic components.

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. s Based on this study's findings and conclusions, recommendations which address these two concerns are provided in the report. The recommendations address:

(1) the establishment of procedures and training of operators to cope with loss of cooling to instrument cabinets, (2) the need to monitor actual conditions (specifically, temperature) in instrument cabinets, and (3) the need for plant technical specification requirements governing the operability of control room cooling and ventilation systems which reflect actual temperature in the instrument cabinets, and (4) the need for specific considerations of this issue in the plant specific evaluation and resolution of the station blackout issue (USI A-44).

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1.0 INTRODUCTION

On June 4, 1984, while both units at the McGuire Station were operating at 100 percent power, a total loss of the control area ventilation system occurred.

Approximately 45 minutes later, the control room received numerous alarms on Unit 1 indicating high temperature (T in reactor coolant loop C. Alarms also were received for pressurizer le6ET)on Unit One 1. hour into the event the ventilation system was declared inoperable, .and I hour after that the control room operators started to reduce power on both units as required by plant technical specifications. During this period other alarms also were received on both units. At approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months into the event, the instrument cabinet doors were opened and cool air from the air-conditioned computer room was moved through the control room doorway to the cabinets by portable fans and ducts.

About 3 hours3.472222e-5 days 8.333333e-4 hours 4.960317e-6 weeks 1.1415e-6 months after the initial ventilation system loss, the control room area ventilation system was returned to service and declared operable. By this time, the control room operators had reduced power on both units to 97 percent.

The extended loss of cooling in the control area led to overheating of several electronic components in the instrument cabinets located in the area.

Overheating of the components caused erroneous signals to be generated by the instrumentation system.

Based on the McGuire event, a study of the problem was initiated and a pre-liminary AE0D report was issued. Based upon this preliminary report, IE

, Information Notice 85-89, " Potential Loss of Solid-State Instrumentation

! Following Failure of Control Room Cooling," was issued on November 19, 1985, to alert licensees of operating nuclear plants of the event and its potential safety concerns. After the IE information notice was issued, three additional events were identified in which the root cause of an instrumentation system

failure was determined to be overheating of electronic components located in cabinets in a controlled environmental area. The additional events occurred at Davis-Besse, Palo Verde 1 and Summer 1 (Refs. 3-6).

A detailed discussion of these four events, the corrective actions taken by i the licensees and the concerns raised by the events are described in the l following sections of this report.

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. e 2.0 OPERATING EXPERIENCE 2.1 Loss of Control Area Ventilation System at the McGuire Station 2.1.1 Event Discussion-At about 8 p.m. on June 4,1984, while McGuire Units 1 and 2 were operating at 100 percent power with the control area ambient temperature at about 75'F, the train B chiller of the control area ventilation system tripped due to low oil level. Prior to this, train A of the system had been removed from service for maintenance. Approximately 20 minutes later the tripped chiller was restarted, but tripped again on low oil level. With all control area cooling lost,-the ambient temperature in the control room started to increase.

Approximately 45 minutes after the initial trip of the control area ventilation i system, the control room received numerous repeated alarms which indicated a high T a in Unit I reactor coolant loop C. Alarms were also received from a Unit l gpessurizer level channel. To prevent further alarms, the affected differential temperature and T circuits were bypassed. There were a few otheralarmsreceivedonbothOUts. (A later review of data logs from the

control room alarm typers revealed no unusual alarms except a loop C high

! T This is possible since not all alarms print on the typers.) The o%fa. tors again attempted to restart the train B chiller at 9 p.m., but were unsuccessful. Are attempt to cross-connect train A chiller to train B air handling unit also failed.

At 9:05 p.m., approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months after the initial trip of train B chiller, the control area ventilation system was declared inoperable. At about 10 p.m.,

the control room and computer room doors were opened, and portable fans and

' ducts were placed in the doorways to move cool air from the air-conditioned

computer room to the control room. The Westinghouse Process Control System (PCS 7300) cabinets, which contain electronic components which are sensitive to

! elevated ambient temperature, were also opened. At 10:05 p.m., one hour after the ventilation system was declared inoperable, the operators started to reduce power on both units as required by the plant technical specifications. At

10: 30 p.m. the chiller was successfully restarted, and at 10:55 p.m. the train l B control area ventilation system was declared operable. At this time, power I reduction on both units was terminated. When normal cooling was restored in i the control area, the malfunctioning instruments returned to normal operation.

I 2.1.2 Analysis and Evaluation of the Event At the McGuire Station, numerous permanent and intermittent failures of printed circuit cards involving reactor trips or spurious instrument indications have been experienced since 1981. The failures appear to be directly attributable to overheating in the PCS 7300 cabinets. The licensee's experience has been '

that, in some cases, the spurious indication (s) disappeared when adequate ventilation was restored to the cabinets. However, generally, failures of the control area ventilation system resulted in erratic signals for over a month i after ventilation was restored.

The McGuire Station Technical Specification Surveillance Requirement 4.7.6 states that the control room temperature cannot exceed 120*F, and if the temperature limit is exceeded, the operating train of the control area 4

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ventilation system is to be declared inoperable. Technical Specification 3.7.6 requires that when one train of the ventilation system is declared inoperable, either the inoperable system must be restored to operable status within 7 days or the plant must be in at least hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in cold shutdown within, the following 30 hours3.472222e-4 days 0.00833 hours 4.960317e-5 weeks 1.1415e-5 months . When both trains of the system are inoperable, Technical Specification 3.0.3 requires that action be initiated within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months to place the unit in hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

The manufacturer, Westinghouse, has recommended that the printed circuit cards not be operated outside a temperature range of 75 F +/- 10 F any longer than is absolutely necessary. Westinghouse has stated that the PCS 7300 equipment specifications call for a normal room ambient temperature of 60*F to 80 F, with a maximum room ambient temperature range of 40*F to 120*F. The early 1970s prototype test showed a 20 F temperature rise in the cabinet over the room ambient. The card performance test was at 140 F (12-hour cycle).

During the event of June 4,1984, the maximum room ambient temperature near the PCS cabinets, as recalled by the operators, was about 90 F, which was well below the the technical specification limit of 120"F. Following the event, the licensee installed temperature monitors at various locations in the PCS cabinet with the highest heat load. With the room ambient temperature near the cabinet at about 72 F, the measured cabinet internal air temperatures ranged from 73 F at the bottom of the cabinet, to a maximum of 109 F at the top of the cabinet.

(The cabinet does not have a fan, and cooling is by natural convection from bottom entry openings in the cabinet to the top exit opening). The thermo-couples, which were placed directly on the instrument racks on which the printed circuit boards are mounted, measured 115 F at the middle rack, to a maximum of 125 F at the top rack. The measured difference between the outside room ambient temperature and the inside instrument rack temperature of the cabinet was greater than expected by the licensee - the expectation probably being based on the prototype testing done by the manufacturer. (It should be noted that the room ambient temperature was 90 F during the event, vice 72 F at which the measurements were made. Therefore, the internal temperatures during the event would have been 15 to 20 degrees higher.) The event occurred during the evening hours of early June when the outside temperature at the McGuire Station site was between 72"F and 76 F. Had the event occurred when the site outside temperature was higher, the control area and the cabinets could have experienced even higher temperatures sooner.

The McGuire Station technical specification surveillance requirement regarding control room ambient temperature states that it cannot exceed 120'F, and this l

limit has never been reached, not even during the 2-1-hour loss of ventilation l

cooling event of June 4, 1984. During that event the maximum room temperature l reached was approximately 90 F, and the maximum temperature experienced by the printed circuit cards was approximately 140 F. As a result, more than one channel of safety-related instrumentation systems experienced problems due to the elevated temperature in the control area. This may have been due to the effect of the elevated temperature on existing degraded components in the instrumentation channels, with the degradation itself probably due to higher-than-normal temperatures in the cabinet.

As stated earlier, the McGuire Station has, prior to this events, experienced numerous failures of solid state electronic components due to elevated control room ambient temperature. The licensee had in the past taken temperature

measurements inside the instrument cabinets of concern, and instituted some corrective actions. However, these corrective actions have apparently not been adequate, as shown by the temperature measurements taken after the June 4, 1984 event.

Following the event, plant maintenance personnel rebalanced the airflow in the control area ventilation system to provide additional cooling to the PCS cabinets. Modifications providing heat sinks on the PCS cards have also since been completed on both units. To prevent unauthorized changes to the settings of the control room area thermostats, covers have been placed over the thermo-stats. The licensee has also developed a procedure (based on correspondence with Westinghouse) to establish forced air cooling of the PCS 7300 cabinets if the control room ambient temperature exceeds 85'F.

Rebalancing of the air flow has apparently already provided some benefits. As stated in the revised LER describing the June 4th event (Ref. 2), in the 5 months prior to rebalancing, 35 card failures had occurred; and in the 5 months after rebalancing, only 13 card failures had occurred. (It should be noted that the majority of these card failures were neither reported nor reportable to the NRC.) The licensee feels that rebalancing the air flow had its greatest impact by increasing the mass flow rate through the cabinets rather than reducing the ambient temperature around the cabinets. However, since rebalancing the air flow around the cabinet, the licensee has not measured temperatures in the cabinet to see the improvement made by the increased mass flow rate through the cabinet.

Based on the history of printed circuit card failures at the McGuire Station, it is apparent that high temperatures in the instrument cabinets would lead to accelerated failure of heat sensitive components in the cabinets. As shown by the measurements taken at McGuire, the temperatures experienced by the electronic components in the cabinets are much higher than the room ambient temperature, and the temperature difference between the inside and the outside of the cabinets during normal operation was higher than the expected 20 F.

Thus, electronic components can be exposed to damaging temperatures (e.g.,

above 140 F) at room ambient temperatures far below the technical specification limit of 120 F.

The corrective actions taken by the licensee following the event of June 4, 1984 (such as rebalancing the air flow in the control room area, modifying the heat sinks on the PCS cards, preventing unauthorized changes to the control room thermostat, and revising procedures to establish forced cooling to cabinets when control room ambient temperature reaches 85*F), should increase the reliability of solid state electronic equipment at the McGuire Station.

However, since the licensee has not measured the cabinet internal temperatures since these modifications have been completed, the actual conditions to which the components are exposed are not known, Specifically, it is not known whether the temperatures experienced by the components in the cabinets are within the specified temperature range when the room ambient temperature is within the range of 60 F to 80 F.

Finally, the McGuire Station has been experiencing printed card failures since 1981, and the licensee was aware that one of the contributing causes of the failures was the ambient temperatures to which the heat sensitive components were exposed. Yet, it was only in 1985 that a comprehensive plan of corrective

action to address the problem was instituted.

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The June 4, 1984 event and other prior experiences at the McGuire Station raise two concerns regarding the effects of high temperature on heat sensitive electronic components in safety-related instrumentation systems. The first effect is the decrease in reliability (i.e., the increase in failure rate) of printed circuit cards and other electronic components due to the high temperature in the cabinets. The second effect is the potential for common cause failure of redundant instrumentation channels due to high temperature caused by a loss of ventilation cooling of redundant instrument cabinets. The two concerns are not necessarily mutually exclusive of each other, as discussed earlier.

Such failures of printed circuit cards in safety-related instrumentation and control systems can lead to malfunctions of control systems, inoperability of instrumentation channels in protection systems, inadvertent actuations and/or inoperabilities of safety features actuation systems including the reactor trip system, and erroneous indications and alarms in the control room.

As mentioned earlier, the corrective actions taken by the licensee should improve the reliability of the instrumentation systems. However, since the licensee has not monitored the actual temperatures in the cabinets since the corrective actions have been implemented, the adequacy of corrective actions cannot be determined. The licensee is not aware of the temperatures which the components of concern are experiencing during normal operation, or what margin exists between normal operating temperatures and design temperature limits.

If a total loss of control area cooling and ventilation system were to occur now, the licensee still does not know when safety-related instruments will begin to respond erratically to the loss of cooling and subsequent high internal cabinet temperatures. '

2.2 Power Supply Failure at Davis-Besse 2.2.1 Event Discussion On September 21, 1982, while Davis-Besse was operating at full power, the control room operators received a full trip alarm of channel 3 of the Steam and Feedwa*er Rupture Control System (SFRCS). Investigation of the trip identified that the field contacts input 48-volt dc/dc power supply had tripped. The power supply was reset, and the input and output voltages were checked and verified to be correct. However, 15 minutes later the power supply tripped again. It was reset a second time, but tripped a few minutes later. After the third trip, the power supply was replaced and the failed unit was sent to the manufacturer for analysis, l This was the first failure at Davis-Besse of a 48-volt dc/dc power supply, but there have been several failures of 24-volt dc and 15-volt de power supplies in the SFRCS, dating back to 1979.

This event was initially reported in LER 82-051 in September 1982. A revision to the LER was issued on January 25, 1986, after the manufacturer had completed an analysis and the licensee had completed the required corrective actions.

It was determined that inadequate design of the SFRCS cabinets had resulted in l overheating of the power supplies in these cabinets; and overheating was a

! primary contributing factor that has led to the failures of the power supplies.

l The revised LER states that by a field change review initiated in 1984, fans i

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were installed in the SFR{S cabinets to proyide additional cooling to the components in the cabinetc. Prior to the restart of the unit in 1986, the licensee plans to ensure the adeluacy of cooling in these cabinets by measuring the temperature in various locations in the cabinets aid of the components.

2.2.2 Analysis and Evaluation of the Event The SFRCS at Davis-Besse is an engineered safety features actuation system (ESFAS) that can sense a loss of main feedwater flow, a break of a main feedwater or steam line,ior a loss of primary coolant forced flow. It provides output actuation signals to isolate main feedwater flow and main steam flow 'and initiate auxiliary feedwater f3cw to the sten gen 6rators. The system also- '

provides trip signals to the turbine ard the anticipatory' reactor trip system.

The SFRCS consists of four logic channels nranged'such that, in general, two' charigefs must trip to provide actuationsof engineered safety features (ESF) equipment. 3t; ,', ,

e During the event, only channel 3 tripped. However, tne potential existed for all four channels to be affected in a \similar manner, if, for example, a common event, such as a loss of ventilation, caused all four instrument channel cabinets to overheat (all SFRCS instrument cabipets are located in the same room). A trip of more than one channel of theLSMCS could have led to isola-tion of the main feedwater and main steam systems, actuation of tte auxiliary feedwater system, isolation of the steam generators,' turbine trip, or reactor

' , trip, dependi,ng are the components affected in the SFRCS cabinets. -

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' Davis-Besse had been experiencing powe/ supply and other component failures in ,

the SFRCS cabinets since 1979, yet,Lonly by lete 1984 yas the licensee able to

( determine the root cause of the fail yes to be overheating of the components.

i' The test program planned by the licensee will assure that t'he modification has been adequate in pro 91 ding the necessar) coolin'g sto the dwponents in the cabinets during normal operation. The'ilcensee will a7so ash re that the ten.peraturet to which the l!ectronic components are exposed are within the specified service conditions of.the equipmentt These corrective actions by the licensee should improve the reliability of the components in the SFRCS cabinets.

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2.3 Failure in ESFAS Cabinet at Palo Verde Unit 1 % \

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2.3.1 Event Discussion i

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On December 16[ 1985, 'at 6:11 p.m., while ,Palo Verde Unit I was operating at 52 -

M percent power, a failure of an electroni: component in the load sequencer

. module for the train A Ensineered Safety Features Actuation System (ESFAS) 4 cabinet occurred. This resulted in several ESF actuations, an inadvertent 4

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operation of the ESF load sequencer, and 'an automatic start of ar. emergency \

diesel generator (EDG). .s t G

e The event occurrdd following the schedbled daily automatic testing of circuits in the train A ESFAS cabinet. The control room operator heard a " popping" sound and' smelled burr}ing such as associated with overheat 6d electronic equip-ment. The source of burned odce was the train A ESFASscabinet., He also noted that all of the lights (in the ESFJIoad sequercer module yere on, and that the I {

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F cooling fan for the load sequencer module was not operating. Since all these conditions were not normal and indicated a malfunctioning load sequencer, an investigation was imediately initiated.

The malfunctioning ESF load sequencer spuriously caused the train A EDG to start, and initiated an invalid load shed and loss of power condition on the train a 4160 volt Class IE bus and its loads. The normal and alternate bus supply breakers were signalled open, and all loads were stripped from the bus.

The problem in the train A ESFAS cabinet also actuated the following ESF signals: Fuel Building Essential Ventilation Actuation Signal (F8EVAS),

Containment Purge Initiation Actuation Signal (CPIAS), and Control Room Essential Filtration Actuation Signal (CREFAS).

Following isolation of the 4160-volt bus and stripping of the loads, the EDG supply breaker did not automatically close as designed, and the bus remained deenergized. To restore power to the bus, the operator attempted to manually close the normal supply breaker. The breaker would not close because of the existing spurious load shed signal from the ESF module. The operator then attempted to manually align the EDG supply breaker in order to energize the bus. This resulted in a trip of the diesel generator and opening 'of the EDG breaker at 6:18 p.m. The indicated cause of the EDG trip was reverse power which is not per design.

Following the EDG trip on reverse power, the EDG continued to run on starting air, at low speed, until it tripped on low lube oil pressure. This mode of operation was also unexpected. This is attributed to the failure of the EDG test circuit in the ESFAS cabinet.

Another attempt to restore power to the bus and itt. loads was made. The normal supply breaker was manually closed after removing its control power fuses.

However, the various load center feeder breakers which supply the loads could not be closed because of the spurious load shed trip signal.

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In order to clear the load shed signal, the train A ESFAS cabinet was downpowered (i.e., deenergized). The load center feeder breakers automatically closed as the load shed signal cleared. This,~in turn, reenergized the ESFAS cabinet and reinitiated the train A loss of power and load shed signals. The normal power supply breaker tripped open again.

l The ESFAS cabinet was downpowered again, and the normal supply breaker of the train A 4160-volt bus was reclosed. To prevent another loss of power and load shed actuation, the train A undervoltage relays were jumpered, and leads of the load shed relays were lifted. The ESFAS functions were then reset and the system was returned to normal operating status. The ESF load sequencer was placed in the manual mode.

To prevent the ESF load sequencer from automatically starting the train A high pressure safety injection (HPSI) pump, low pressure safety injection (LPSI) pump, and containment spray pumps, the control fusesTto these pumps were removed prior to downpowering the ESFAS cabinet. This rendered these pumps

8 inoperable. During the initial phase of the event, at 6:12 p.m., the train B essential chiller had tripped on low refrigerant level, rendering all train B equipment inoperable.

Control room operators had recogniz'ed that the loss of both trains of HPSI, LPSI and containment spray systems required entry into plant Technical Specification 3.0.3, which required that actions be initiated to bring the plant to hot shutdown within I hour. Due to the various activities in the control room, entry into the technical specification action statement was delayed by 38 minutes; shutdown-of the reactor was initiated at 7:50 p.m.,

and the unit achieved the required hot shutdown condition at 00:46 a.m. on December 17, 1985.

On December 17, after the ESFAS cabinet had cooled, test.ing of the ESF load sequencer revealed proper operation of all functions except the auto-test function test point for the EDG. The load sequencer locked-up during its attempt to perform the auto-test.

2.3.2 Analysis and Evaluation of the Event The cause of the event has been traced by the licensee to the failure of a fan in the ESF cabinet. Failure of the fan allowed the ESF load sequencer module to overheat and malfunction. As immediate corrective actions, the ESF load sequencer was replaced and the cabinet door was temporarily removed. The

, failed fan was replaced with two larger capacity fans. A control room alarm, that annunciates on high cabinet exit temperature has since been installed.

These corrective actions have been implemented on all three Palo Verde units.

The licensee is planning to mea _sure the actual teraperatures in the cabinets to monitor the-effectiveness of the corrective actions.

The failure of the EDG breaker to automatically close and load the train A bus was later determined by the licensee to be the result of a component failure in the ESFAS cabinet. The component failure apparently left the EDG in the " test" mode rather than transfer it to the " emergency" mode. (Automatic closure of the EDG breaker on a loss of power signal is only designed to occur when the EDG is operatinc in the " emergency" mode.)

The two ESFAS cabinets (train A and train B cabinets) for each unit of the Palo Verde station are located side by side in the instrument cabinet area near the control room panels. The ESFAS environs are cooled by the control room ventilation system. Each cabinet contains the electronic modules and equipment necessary to perform the ESF functions of one train, such as load sequencing, load shedding, loss of power, EDG starting and actuating FBEVAS, CREFAS and CPIAS.

Although the event resulted in total loss of all ac power to one train of safety-related equipment and malfunction of one train of ESFAS, the plant did not experience any transient that necessitated a unit trip. However, cross-train actuation of the CREFAS initiated the train B essential chiller, which subsequently tripped. This rendered train B equipment inoperable. In the process of correcting the spurious actuation of train A ESFAS, the operators had to disable the train A HPSI, LPSI, and containment spray pumps. Thus for a short period, both trains of HPSI, LPSI and containment spray systems were inoperable, requiring the plant to be brought to a hot shutdown condition per

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plant technical specifications. The control room operators were placed in a stressful situation and had to take actions without specific procedures involving: pulling fuses, jumpering relays, lifting leads and manually taking ,

control of equipment automatically actuated by the ESFAS, in order to cope with the_ problem. The operators were so involved and busy in controlling the spurious actuation of the ESFAS that they delayed (by 38 minutes) entry into a required technical specification action statement. The operators performed creditably in recognizing the spurious actuations of the ESFAS equipment and in taking the actions they did to safely control the plant.

Although only one cabinet containing electronic components was affected during this event, the potential existed for others being affected in both the ESFAS '

and the Plant' Protection System (PPS) since, for a short period of time, both trains of essential chillers were not operating and the control room area cooling was not available. Had the operators been delayed in restoring power to the train A bus (and thus in returning Chiller A back to service), multiple trains of electronic equipment could have been affected by the increasing ambient temperatures in the control room and cabinet area.

The corrective actions taken by the licensee should adequately address the problem of loss of forced cooling in a particular ESFAS cabinet. The licensee's plan of measuring actual temperatures in the cabinets will verify the adequacy of forced cooling provided to the electronic components during operation.

This event emphasizes the concern regarding the potential for common cause failure of instrument channels due to high temperatures caused by loss of cooling to the cabinets which house sensitive electronic components. Failure of electronic components, in this case, led to the total loss of power to one train of essential safety-related equipment, and at the same time prevented access to all sources of offsite and onsite power supplies. Due to subsequent failure of redundant equipment, the event led to a total loss of several ESF systems. Fortuitously, the need for these systems did not arise during the event, and the plant was operated safely throughout the event by the operators.

The event also illustrates 'the point that failure of electronic equipment, components or circuits can be in any of several modes; in this case, the sequencer module failed in such a mode as to lock the sequencer in a state that corresponded to a particular 1-second period in the logic sequence. Such a failure mode would not likely be predictable. Furthermore, procedures to cover the myriad failure modes of all electronic components or circuits is not

-practical. Thus, it is not expected nor desirable that plants have specific procedures covering all failure possibilities of safety-related instrument systems due to electronic component failures. However, operators need to be

. made aware of (and be provided generalized procedures or training for) the type of problems that could occur should there be extended loss of cooling of electronic components in instrumentation and control system cabinets.

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2.4 Operational Experience at Suniner 1 2.4.1 Event Discussion

, On December 6, 1982, with the plant operating at 30 percent power, spurious low and high indications were seen on the main control room indicator for steam generator C steam flow (FI-494). The cause of the occurrence wes attributed to +

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a component breakdown in a circuit board in the instrument loop when exposed to i the normal cabinet ambient conditions. The licensee concluded that electronic components on this circuit board had deteriorated and experienced intermittent failure even though the licensee believed that the cabinet internal temperatures were within their design limits. The defective card was replaced and the instrument loop was returned.to service; no other corrective actions were taken.

i On December 16, 1982, while the unit was operating at 50 percent power, one channel of the main control board pressure indicator (PI-951) for reactor building pressure failed low. The inoperable channel was placed in a tripped condition in accordance with plant technical specification requirements.

The cause of the event was due to the failure of the loop power supply card.

When the door of the cabinet bay, in the relay room, where the component was located, was left open, the channel indication returned to normal. The failure of the card was attributed to elevated ambient conditions in the cabinet.

Cabinet cooling was provided by convective heat transfer through natural cir-

! culation of air in the cabinets. The licensee concluded that due to the density of the cards and location of the power supply in the cabinets, insufficient cooling had created conditions leading to card overheating and recurring instrument loop failures. The licensee initiated plans at that time to install additional air conditioning in the relay room in which the

instrumentation system cabinets are located, and started an engineering evaluation to determine any additional measures.

On January 8, 1983, the main control board indicator for steam generator A feedwater flow (FI-477) failed low. The failure was attributed to the breakdown of circuit board electronic components when exposed to internal i cabinet temperature. Immediate action was initiated to lower instrument cabinet and room temperature to prevent recurrence. Testing by the licensee i indicated air stagnation in the relay room from inadequate ventilation system 4

distribution. Hence, as a long term corrective action, the licensee planned to I alter the relay room ventilation flow path to provide more effective cooling to the instrument cabinets. Based on this, the licensee cancelled the previous plan of installing additional air-conditioners for the room. The corrective action was implemented in mid-1983, and apparently some tests were performed by the licensee to verify the efficacy of the modification.

4 On February 10, 1983, with the unit at 50 percent power, the A feedwater flow indicator FI-477 failed low again. The power supp1" circuit board was found to have failed. The circuit card was replaced ano the instrumentation channel was returned to service; no additional corrective action was taken. ,

On June 6,1983, a feedwater flow channel failed due to a power supply card failure. The card was replaced and the channel was returned to service.

On September 23, 1983, while at full power, a signal comparator associated with a channel of containment pressure failed and the defective bistable card was replaced.

4 On November 24, 1983, with the plant in Mode 3, the pressurizer pressure trip j bistable PB-455D began to alternately trip and reset. The problem was caused i by a defective electronic circuit board which was replaced.

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On February 7, 1984, the reactor tripped while operating at 100 percent power because of a low-low level in steam generator B (Ref. 7). This was due to the partial closing of a main feedwater regulating valve when an electronic card in its control circuitry failed.

The licensee believes that the modification of the ventilation system has corrected the problem of overheating of components in the instrument cabinets located in the relay room. Since the implementation of the modification, the licensee has not experienced any failure of electronic components due to elevated temperature. (Some failures that occurred after the modification were attributed to deterioration of circuit boards and components due to the earlier exposure to elevated temperature.)

2.4.2 Analysis and Evaluation of Operational Experience The relay room at Summer 1 contains the instrumentation cabinets, including the PCS cabinets, and is cooled by a heating, ventilation and air-conditioning (HVAC) system that is separate from the control room cooling system. With this HVAC functioning normally and with the relay room ambient temperature at its normal operating temperature (nominally 75*F +/- 10 F), the licensee was experiencing electronic component failures in the PCS cabinets and other instrument cabinets located in the relay room. (Like other licensees, this licensee also took a while to finally identify that elevated internal cabinet temperature was the root cause of some of the failures of electronic components that the plant had been experiencing.) The failures of the components have led to the inoperability and trips of plant protection, control and indicating channels. In some cases, the failures caused a transient that led to a reactor trip. In all cases, the failures occurred randomly; i.e., multiple failures due to loss of cooling have not been experienced at Summer 1.

At Summer 1, the licensee has seen a definite improvement in the reliability of electronic components in the cabinets in the relay room since the adjustments to the HVAC system in 1983. The licensee has measured the ambient temperature in the relay room at different locations after the HVAC system adjustments and, based on the room ambient temperature being within the design limits of the ,

instrumentation system equipment, considers the corrective action taken as adequate. Hcwever, the licensee apparently has not ascertained the tempera-tures in the cabinets to assure that the components themselves are not exposed to higher than the design temperature limits.

2.5 Other Events Two other events are presented here as further examples:

At San Onofre 2 on May 16, 1983, while the unit was operating at 80 percent power, the operator received a core protection calculator (CPC) failed indication and alarm which was caused by the failure of channel D of the CPC. The failure of CPC channel D was attributed to a degraded power supply which was probably caused by the loss of a CPC cabinet cooling fan (due to bearing failure). The power supply and cooling fan were replaced and the CPC channel D was declared operable (Ref 8). ,

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At Waterford Steam Electric Station Unit 3 on January 22, 1986, while the unit was operating at 100 percent power, a control element assembly (CEA) inadvertently dropped into the core, causing a CPC reactor trip. Investiga-tion into the reason for the event revealed that the fan blower toggle switch for the cabinet cooler was in the off position (possibly due to personnel bumping into the switch). This may have overheated the CEA circuitry, resulting in the dropped rod (Ref. 9).

In both the above cases, the failed fans and components were replaced and the system was returned to operation. How long the fans were off or what effect the elevated temperature in the cabinet may have had on other components was not determined.

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3.0 GENERIC APPLICABILITY OF THE PROBLEM OF HIGH AMBIENT TEMPERATURE Since heat sensitive electronic components are present in the instrumentation and control systems at many operating nuclear plants, the problem of elevated cabinet internal. temperature and the consequential failures of electronic components are likely to be experienced at many plants. However, the cause of these failures may not be immediately recognized. A search of the Sequence Coding and Search System (SCSS) data base for failures or events involving high l ambient temperatures was not successful in identifying any more events. Hence, '

a limited search of the SCSS was conducted for events at operating Westinghouse i units with Solid State Protection Systems (SSPS) that involved the following types of ambient or cabinet internal temperature-related problems seen at the McGuire and Summer stations:

o Instrument loop power supply card failure, o Signal isolator card failure, o Lead / lag amplifier card failure, o Signal comparator card failure, o Universal logic card failure, and o Controller-driver card failure.

The events thus obtained are listed the appendix to this report. A review of these events found that the types and characteristics of card failures experienced by the McGuire and Summer units are also evident at the Salem, Cook, North Anna, Trojan, Beaver Valley, Sequoyah and Farley units. In most of the events, the licensees did not determine the root cause of failure, and none of the licensees identified inadequate instrument cabinet cooling, or elevated cabinet internal temperature, as a possible root cause of the failures.

Many of the failures are treated as random failures, and the corrective action generally has been to replace the failed component and return the inoperable instrumentation or control system to operation. One possible reason for discounting high ambient temperature may be that these instrument cabinets are

< located in a controlled " mild" environment and the plants have not experienced an extended loss of HVAC system affecting the area in which these cabinets are located. However, as seen at McGuire and Summer, the temperatures experienced by the electronic components in the cabinets can be much higher than the room ambient temperature in which these cabinets are located. Depending on the location of the cabinet, its heat load, the external and internal ventilation air flow of the cabinet, the arrangement of the components in the cabinet, the method of cabinet cooling (whether the cabinet is cooled by natural convection or forced cooling), etc., the temperature at the component could be above its design limit (normally 140 F). This could be the case even when the ambient room temperature is within specified limits (normally 75 F +/- 10'F). Since internal temperatures are unique to each cabinet, unless actual measurements are taken, there is no assurance that the components are operating within their design service temperature limits.

Since these instrument cabinets are normally located in areas with a " mild" controlled environment, regulatory requirements regarding equipment environ-mental qualification generally have not been applied to the cabinets or the components in most operating reactor units. Hence considerations of aging, operation under normal and accident conditions, ongoing qualification, etc.,

have, in general, not been considered for these cabinets and components.

It should be noted that even at the plants where elevated ambient temperature and/or inadequate cooling was eventually established as the principal root cause of many of the electronic component failures (e.g., McGuire, Summer and Davis-Besse), several years of review were involved before the root cause determined and corrective action programs established.

As far as we have been able to determine, only those licensees who have determined that elevated cabinet temperatures due to inadequate cooling was a cause of electronic component failures, have measured or are planning to measure temperatures in the cabinets of concern. Even in these cases, only the cabinets of concern are investigated (i.e., other cabinets in the same area are not being monitored or tested).

Although the SCSS search was confined to Westinghouse PWR units, based on those events and on the events at Palo Verde and Davis-Besse, we believe that the problem is potentially generic to all operating reactors that utilize heat sensitive electronic components in the plant instrumentation and control system. The printed circuit cards and other electronic modules used in safety-related instrumentation and control systems in PWRs and BWRs are very similar and are subject to similar failure modes. They are, in general, designed to be installed and operated in controlled mild environments subject to similar design specifications.

Based on operational experiences it can be concluded that operation of heat sensitive electronic components in inadequately cooled cabinets will lead to increased failure rate of printed circuit cards and other heat sensitive components located in the cabinets. The potential for common cause failure of redundant instrumentation channels due to inadequacy or loss of cooling to all the cabinets also exists at those plants where redundant equipment is located in a common area or cooled by a common HVAC system. If during normal operation, the components are operating in a higher-than-normal cabinet internal temperature, deterioration of heat sensitive components will occur, which, in turn, will tend to leave the components more susceptible to common cause failures.

Failures of such components in safety-related instrumentation channels have led to malfunctioning of control systems, inoperability of protection channels, inadvertent actuation of safety features actuation systems including reactor trip, loss of one train of ESFAS, and erroneous indications and alarms in the control room.

Based on the types of failures experienced, the potential exists for redundant channels to be affected when a sustained loss of cooling to multiple cabinets occurs (e.g., loss of the HVAC system, loss of power, station blackout), causing problems in more than one channel or train of safety-related systems.

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Continuous or routine monitoring of the actual temperatures of each cabinet that contains heat sensitive electronic components is one way of assuring that the components are always operating within their design limits.

Most operating nuclear plants are subject to technical specification requirements, similar to that at the McGuire Station, governing the operability of the control room cooling system. The specifications, in general, require that a shutdown be commenced within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months following the loss of the control room cooling system (i.e., inoperability of both trains of the cooling system) and that the unit be placed in hot standby within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . (In the case of loss of forced cooling to a particular instrument cabinet, in general, there is no technical specification requirement specifically addressing such a situation.) The bases for the technical specification requirements for the operability of the control room cooling system are to ensure that (1) the room ambient air temperature does not exceed the allowable temperature for continuous duty rating for the equipment and instrumentation cooled by this system, and (2) the control room will remain habitable for operation personnel during and following all credible accident conditions.

As shown by the events at McGuire, Palo Verde and Davis-Besse, instrumentation systems containing heat sensitive components could start to malfunction well within I hour after a loss of cooling and at control room temperatures substantially below 120 F. This suggests that the time period and the temperature limit allowed by plant technical specifications to accommodate control room ventilation and cooling systems have not conservatively considered the effects of elevated temperatures on electronic components located in the area cooled by the system. Furthermore, the time periods currently allowed by the plant technical specification could give the control room operators a false sense of security since they would assume that in the time frame permitted by the technical specifications, the plant would still be capable of being safely shutdown.

At the McGuire Station (where the operators had experienced several earlier losses of control room cooling prior to June 4,1984, and where the operators were aware of the effects of high ambient temperature on instrument components in the PCS cabinets) it took 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to provide alternate cooling to the cabinets. As one of the corrective actions, the licensee revised plant procedures to establish forced cooling to cabinets in the control area when control room temperature reaches 85 F. Operators at plants that have not experienced a sustained loss of cooling to instrument cabinets could have more difficulties in coping with such an event.

It would appear that there is a need for the operators to be alerted and trained, and plant procedures established, for such events. Plant procedures dealing with loss of HVAC systems or loss of cooling to cabinets, in general, do not consider the consequences of malfunctions of redundant instrumentation channels. Plant procedures which require opening of cabinet doors and establishing alternate cooling to cabinets could be vitally important in coping with such events.

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4.0 STATION BLACK 0UT AND HIGH AMBIENT TEMPERATURES A station blackout at a nuclear plant will lead to the loss of all heating, ventilation and air-conditioning systems at the plant. In the ongoing evaluation of total loss of all ac power transients (USI A-44, Station Blackout), an item of consideration is the unavailability of normal heating, ventilation and air-conditioning systems in the plant. The evaluation, as documented in NUREG-1032 (Ref. 10), states that equipment needed to operate during a station blackout and equipment required for recovery from a station blackout would have to maintain operability in environmental conditions (e.g.,

temperature, pressure, humidity) that could occur as a result of the blackout.

Otherwise, failures of necessary equipment could lead to a loss of core cooling and decay heat removal during the event or failure to recover from the event upon recovery of ac power. Instrumentation and control elements of components required during station blackout are the most likely to be impacted by adverse environments. This is evident from the event at McGuire where several channels of instrumentation were adversely affected by high ambient temperature, and from the event at Palo Verde where loss of cooling to an instrumentation system cabinet caused a total loss of all ac power to one train of safety-related equipment and at the same time prevented recovery of available offsite and onsite ac power.

These considerations have, to a certain extent, been taken into account in the staff's resolution of USI A-44. In the draft regulatory guide on station blackout (Ref.11), which is part of the proposed technical resolution of the issue, the evaluation of plant-specific station blackout capability includes the following:

The design adequacy and capability of equipment needed to function in environmental conditions associated with a station blackout should be evaluated. All ac-independent decay heat remove.1 systems and associated equipment needed to function during a station blackout should meet design and performance standards that ensure adequate reliability and operability in extreme environments that may be associated with a station blackout, including hazards due to severe weather. Work that has already been performed need not be duplicated. For example, if safety-related equipment needed during a total loss of ac power has been qualified to operate during environmental conditions associated with a station blackcut (e.g., without heating, ventilating and air-conditioning systems operating), additional analyses need not be performed.

Operating experience discussed earlier in this report has shown that these considerations should not only apply to " front-line" mitigation equipment needed to function during a station blackout but also to support or dependent equipment. Specifically, the operability of instrumentation and control system equipment, whose malfunction would degrade or disable needed front line equipment, must also be assured. The event at Palo Verde is a prime example.

The malfunction of the ESFAS load sequencer due to the loss of cabinet cooling caused a load shedding and lockout of all safety-related loads on the safety-related buses. If such a spurious actuation were to occur during a station blackout (and it might occur on both trains of a safety-related power system), it could prevent recovery from the blackout event. Furthermore, the event at Palo Verde points out that a loss of instrument cabinet cooling has the potential for both initiating a station blackout event, and preventing

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recovery from the station blackout. Similarly, the malfunction or erratic response of steam generator level instrumentation or steam generator rupture control systems, such those at Summer or Davis-Besse, can prevent the proper functioning of the auxiliary feedwater system during a station blackout.

In addition, the USI A-44 statement regarding equipment qualification must be carefully applied to avoid oversights or omissions. For example, the PCS cabinet at McGuire was qualified to operate in an ambient temperature of 120 F; yet, because of plant-specific conditions, the instrumentation started to malfunction at a room ambient temperature of about 90*F. Hence, in the evaluation of plant specific station blackout capability, licensees should not rely solely on equipment qualification work that does not consider plant specific conditions (e.g., as built locations, actual heat loads, ventilation system distribution).

NUREG-1032 states that for equipment like instrument cabinets, located in control rooms and auxiliary buildings, opening doors should allow enough heat to be dissipated to maintain equipment in an acceptable operating environment for the 4-hour or 8-hour duration of the event. This suggestion is also endorsed in ANS 58.12, Draft Standard for Evaluation of Response Capability for Loss of All AC Power at Nuclear Power PTants, which liTely will be endorsed in an NRC regulatory guRe. The events at McGuire and Palo Verde, however, show that simply opening doors may not be adequate. The operators at McGuire had to take several measures, including opening cabinet doors, as well as computer room and control room doors in addition to providing fans with ducts to transfer air-conditioned cool air from the computer room to the cabinet.

Based on these considerations, we believe that in the evaluation and resolution of USI A-44, more detailed attention must be given to the effects of environmental conditions on instrumentation and control system elements, especially heat sensitive electronic components. In particular, the actual temperature and condition of components located inside instrumentation system cabinets and other enclosures should be considered in plant-specific evaluations.

5.0 FINDINGS AND CONCLUSIONS i- Based on the review and evaluation of events involving instrumentation and control system failures due to loss of or inadequate cooling to heat sensitive 4 electronic components as discussed in the previous sections, the following

-findings and conclusions are presented:

5.1 McGuire Station Eve _

, (1) The total loss of control area ventilation system cooling at the McGuire Station on June 4, 1984 caused some safety-related ,

instrumentation channels on both units to be adversely affected and to j behave erratically due to the increasing ambient temperature.

(2) During the McGuire event o >erators had to 3rovide alternate cooling to the affected instrument ca)inets to keep t te instrumentation systems operable.

(3) The degradation of the instrument system components occurred even though the room ambient temperature was well below the equipment design service and technical specification limits.

The maximum room ambient temperature in the control room area during the event was approximately 90 F. The technical specification limit for the room is 120 F. The process control system (PCS 7300) equipment i

specification requirement for maximum ambient temperature is also 120 F. However, during the event when the control room ambient temperature was in the 85"F to 90 F range, certain safety-related instrument channels in the PCS cabinets were adversely affected.

(4) At McGuire Station, operation of the PCS cabinets within the specified i room ambient tem)erature limits still resulted in internal temperatures hig1er than predicted by the manufacturer.

, The manufacturer of the PCS has recommended that the printed circuit cards and other electronic comporents in the cabinets not be normally l operated routinely outside a range of 75*F +/- 10 F. The equipment

specification calls for normal room ambient temperature of 60*F to 80 F,

and maximum room ambient temperature range of 40*F to 120*F. Prototy-4 pical testing showed a temperature rise of 20'F in the cabinet over j the room ambient. The card performance test was at 140*F (12-hour cycle).

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! Following the event, the licensee measured the internal temperatures in

one of the PCS cabinets at McGuire Unit 1, and found that with a room ambient temperature of 72*F, the temperatures ranged from 73'F to 109'F in the air space in the cabinet. The temperatures on the instrument racks on which the printed circuit boards are mounted measured 115'F to 125*F. The temperature increases in the cabinet and in the area adjacent to the components were higher than the expected 20*F rise.

,! (5) The plant technical specification requirements for the McGuire Station i regarding ambient temperature limits and times required for actions were inadequate and have not fully considered the effects of the actual

! temperature rises in the cabinets during a loss of cooling to the

! instrument cabinets and their components.

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i Technical specifications at the McGuire Station require actions to be_ initiated 1-hour following the total loss of control area 1 cooling to place the affected units in the hot standby mode within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . During the event, certain safety-related instrumentation channels of one unit began to be adversely affected 45 minutes after

-loss of tne cooling system.

(6) At plants which have implemented corrective measures to improve air cooling to vital electronic components, the efficacy of these corrective measures for a loss of control area cooling has not been adequately determined.

Corrective actions taken at the McGuire Station (such as rebalancing the air flow in the control area, modifying the heat sinks on the PCS cards, preventing unauthorized changes to the control room thermostat, and revising procedures to establish forced cooling to cabinets when

the control room ambient temperature reaches 85'F), reduce the

temperatures in the electronics cabinets which should increase the i reliability of electronic equipment at the plant. However, since

implementing the corrective actions, the licensee has not measured the cabinet and component temperatures. Further, the licensee is not aware of the temperatures which the components are now experiencing during s- normal operation, or what margin exists between normal operating tem-

peratures and design temperature limits. If a total loss of cooling and ventilation were to occur, the licensee still does not know when safety-related instrumentation might be expected to begin to respond

. erratically to the loss of cooling and subsequent high internal cabinet temperatures, i

(7) The significant contribution of the adverse effects of high internal cabinet tem)erature on electronic components, has not been readily recognized )y plant personnel.

The McGuire Station has been experiencing printed card failures since 1981, and the licensee was aware since then that elevated ambient temperature was one of the contributing causes. Yet, only in 1985 was

. a comprehensive plan of corrective actions instituted to address the problem.

! 5.2 Davis-Besse Event The significant contribution of the adverse effects of high internal SFRCS cabinet temperature on electronic components, had not been readily or fully recognized by plant personnel.

Davis-Besse plant has experienced recurrent power supply failure in the

SFRCS cabinets since 1979, yet, only in 1984 was the root cause of the problem determined to be overheating of the components.

5.3 Palo Verde 1 Event l

! A loss of instrument cabinet cooling has the potential for both initia-l ting a loss of all ac power to safety-related electric buses and

preventing recovery from available ac power sources.

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'At Palo Verde 1 the failure of the cooling fan in an ESFAS cabinet .

caused the malfunction of the ESF load sequencer module in the cabinet t due to overheating. This caused the start of an emergency diesel ,

generator, and initiated a load shed and a total loss of power to one train of the safety-related electric power system. The malfunction also prevented all automatic and manual operation of the associated breakers and prevented recovery of power to'the train. Only after disconnecting relay leads, attaching jumpers, pulling fuses and manually closing breakers was recovery possible.

5.4 Generic Aspects of the Problem

(1) The significance of the adverse effects of elevated ambient temperature F on electronic components has not been fully recognized by operations personnel at operating nuclear plants, r

i Operating Uet,tinghouss PWRs have experienced failures which are similar

, to the t:mes experienced by printed circuit cards and other heat sensitive electronic components at the McGuire and Summer plants. In most of these cases, the licensees had not determined the root cause of failure. The licensees had not investigated the possibility that j inadequate instrument cabinet cooling or elevated ambient temperature j is a possible root cause of failure.

1 (2) The potential for high temperatures adversely affecting solid state electronic components in safety-related instrumentation and control systems, is generic to all operating PWRs and BWRs.

l (3) The time period and temperature limits allowed by the technical l specifications for the loss of control area cooling system, have not

! adequately considered the actual temperature rise in the cabinets.

The technical specification requirements regarding control room cooling

and ventilation systems are, in general, similar to those at the McGuire Station. Technical specifications at the McGuire Station had a limit of 120 F for room ambient temperature (i.e., the operating i cooling system is considered operable when control room ambient

! temperature is less than or equal to 120*F), and require actions to be

initiated I hour following the total loss of control room area cooling

(i.e., inoperability of both trains of the cooling system). Such

allowances in the technical specifications tend to give the operators

a false sense of confidence, since they would assume that within the i time periods and temperature limits provided, the plant would continue j to operate safety.

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L ~ _ _ _ _. ____ ..._ _ _ _ _ _ _ __ _.,_ _ _ _ ___. _ ._ _ ._______ _ , _ _

s s (4) The effectiveness and efficiency of the control room operators' response to a loss of control area cooling or loss instrument cabinet cooling is dependent on prior experience with such events.

The operators at the McGuire Station had experienced previous loss of cooling events. As a result, they anticipated the effects on the instrumentation system and were able to provide alternate cooling to the cabinets. The operators at Palo Verde 1, who had not had previous exposure to a similar event, had more difficulty in coping with the event.

(5) Operation of heat sensitative electronic components in inadequately cooled cabinets or in elevated environmental temperatures can lead to increased failure rate of the components.

(6) The potential for common cause failure of redundant instrumentation channels exists at those plants where redundant equipment is located in a common area or cooled by a common HVAC system.

(7) Failures of instrument system components due to overheating have caused malfunction of control systems, inoperability and spurious actuation of protection and ESFAS channels, inadvertent actuation and failure of a train of ESFAS, and erroneous indications and alarms in the control room.

5.5 Station Blackout (1) In the proposed resulution of the station blackout issue, environmental considerations following a loss of normal HVAC during a station blackout will not be applied to instrumentation and control system equipment whose malfunction could impact the operability of equipments and systems needed during the event and for recovery from the event.

In the proposed resolution of the issue, environmental considerations apparently only apply to ac-independent decay heat removal systems and associated equipment needed to function during a station blackout.

Hence, eouipment like the ESF sequencer at Palo Verde 1, which malfunctioned due to high environmental temperature, would not be considered. However, such equipment is needed for recovery from a station blackout.

(2) Plant-specific equipment qualification considerations, such as as-built locations, actual heat loads, and HVAC system distribution, may not be adequately considered for the environmental conditions associated with a station blackout.

(3) Operational experience shows that for instrument cabinets located in control rooms and auxiliary buildings, o)ening doors to allow heat may not maintain ecuipment in an accepta)le tem)erature environment for the 4-hour or E-hour duration of a station )lackout.

6.0 RECOMMENDATIONS The following recommendations aimed at addressing the two concerns raised by j the problem of high temperature and its effects on heat sensitive electronic components (i.e., decreased reliability of the components and the potential for coninon cause failure of redundant components) in nuclear power plant instrumentation and control systems, are forwarded to NRR for consideration.

(1) Procedures for (a) loss of HVAC systems supplyino instrumentation and controlsystemequipmentroomsandareas,and(b)lossofforced cooling to instrument cabinets, should be provided. Such procedures should include directions for providing alternate cooling to instrument cabinets affected. The procedures should also include corrective actions to be taken in the event that erratic responses and/or failures of instruments in the plant's protection, control and indication systems occur due to increasing temperatures. Alarms and/or indications should be provided to alert operators of the loss of cooling.

(2) Training of control room and plant operators in using the procedures should be provided.

(3) Supplemental cooling equipment should be readily available and identified for use in the event.

(4) All operating nuclear plants that utilize heat sensitive solid-state electronic comporents in the plant instrumentation and control systems, should periodically measure or continuously monitor the environmental conditions inside the instrument cabinets that contain these components to determine if the components are operating within their design limits.

Continuous temperature monitoring during normal plant operation is preferable, and can be used to alert operators of impending problems, although, periodic measuring is also considered acceptable. As a minimum, a once per refuel cycle test of all such cabinets to obtain inside temperature measurements is considered necessary. The temperature measurements are needed to confirm the design margin of the electronic components in the cabinets and to periodically verify that there is no undetected cooling degradation in the cabinets.

(5) The room ambient temperature limit specified in the plant technical specifications for operability of the control room cooling and ventilation systems, should reflect the actual measured temperatures in the safety-related instrumentation and control system cabinets located in the control room area.

(6) In the ongoing plant-specific evaluations associated with the resolution of USI A-44, Station Blackout, the following considerations regarding the effects of high ambient temperature on solid-state electronic components should be included:

(a) The design adequacy should be evaluated for instrumentation and control system equipment needed to function during and recovering from a station blackout, as well as other equipment whose malfunc-l tion would impact operability of such equipment.

l

(b) Plant-specific equipment qualification data should be required unless the equipment qualification data can be verified by actual measurements of as-built and as-installed conditions.

In recognition of the common NRC and nuclear industry objective of resolving identified safety issues relating to the operation of nuclear plants in the most effective manner, those recommendations above dealing with procedures, training and hardware modifications could be turned over to an appropriate industry group for consideration and followup action.

REFERENCES

1. Duke Power Company, LER 84-018, Docket No. 50-369, July 3, 1984.

2. Duke Power Company, LER 84-018 Rev. 1, Docket No. 50-369, March 22, 1985.

3. Toledo Edison Company, LER 82-051, Docket No. 50-346, October 20, 1982.

4. Toledo Edison Company LER 82-051, Rev. 2, Docket No. 50-346, February 10, 1986.

5. Arizona Nuclear Power Project, LER 85-083, Docket No. 50-528, January 15, 1986.

6. South Carolina Electric and Gas Company, LER 82-016 Rev. 2, Docket No. 50-395, January 13, 1983.

7. South Carolina Electric and Gas Company, LER 83-003, Docket No. 50-395, February 7, 1983.

8. Southern California Edison Company, LER 83-056, Docket No. 50-361, June 15, 1983.

9. Louisiana Power and Light, LER 86-002, Docket No. 50-382, February 21, 1986.

10. U.S. Nuclear Regulatory Commission, " Evaluation of Station Blackout Accidents at Nuclear Power Plants - Technical Findings Related to Unresolved Safety Issue A-44," NUREG-1032, May 1985.

11. U.S. Nuclear Regulatory Commission, " Draft Regulatory Guide and Value/ Impact Statement - Station Blackout," Office of Nuclear Regulatory Research, SI 501-4, March 1986.

-,,,As> L. _. p. 1,. s &as, e..o s

-i- e.

1 APPENDIX EVENTS INVOLVING FAILURES IN WESTINGHOUSE SOLID STATE PROTECTION AND PROCESS CONTROL SYSTEMS l

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FORM 1 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 272 1982 031 0 8206100244 173966 05/02/82 DOCKET:272 SALEM 1 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG ABSTRACT WHILE TESTING VALVE 11SW122 FOLLOWING MAINTENANCE, A NO. lA SAFEGUARDS EQUIPMENT CONTROL (SEC) CABINET ACTUATION OCCURRED. N0. IA DIESEL GENERATOR STARTED AND ALL SAFETY RELATED LOADS, WITH THE EXCEPTION OF NO. 11 SAFETY INJECTION (SI) PUMP WAS DECLARED INOPERABLE, AND ACTION STATEMENT 3.5.2.A WAS ENTERED. THIS OCCURRENCE CONSTITUTED OPERATION IN A DEGRADED MODE IN ACCORDANCE WITH TECH SPECS 6.9.1.9.B. SEE:

82-017,81-119, 81-085,81-083, 81-014. THE SUPPLY BREAKER TO NO.11 SAFETY INJECTION PUMP LOCKED OUT ON THE ANTI-PUMP FEATURE DUE TO A SPURIOUS SEC CABINET ACTUATION; THE SEC SIGNAL WAS DUE TO A FAILED CARD FILE ASSEMBLY IN CONJUNCTION WITH NOISE FROM THE VALVE 11SW122 CLOSE SOLEN 0ID. THE BREAKER LOCK OUT WAS ATTRIBUTED TO THE SPURIOUS NATURE OF THE SEC ACTUATION. N0. 11 SI PUMP WAS DECLARED OPERABLE, AND ACTION STATEMENT 3.4.2.A WAS TERMINATED.

FORM 2 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 272 1982 075 0 8211020092 177960 09/27/82 DOCKET:272 SALEM 1 TYPE:PWR REGION: I NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS C0.

SYMBOL: PEG ABSTRACT THE CONTROL ROOM OPERATOR DISCOVERED THAT PRESSURIZER LEVEL CHANNEL III WAS READING HIGHER THAN THE OTHER CHANNELS. CHANNEL III WAS DECLARED INOPERABLE, AND LIMITING CONDITION FOR OPERATION 3.3.1.1 ACTION 7 WAS ENTERED. THE CHANNEL BISTABLE WAS PLACED IN THE TRIPPED CONDITION. THE EVENT CONSTITUTED OPERATION IN A DEGRADED MODE IN ACCORDANCE WITH TECH SPEC 6.9.1.9.B. INVESTIGATION INTO THE CAUSE OF THIS FAILURE AND A PREVIOUS ONE ON AUGUST 12, 1982 IS BEING CONDUCTED. A SUPPLEMENTAL REPORT WILL BE SUBMITTED UPON COMPLETION OF THE INVESTIGATION AND REPAIR OF THE PROBLEM.

FORM 3 LER SCSS DATA 07-22-E6 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 272 1983 046 0 8311030118 186445 09/24/83 DOCKET:272 SALEM 1 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG r FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG ABSTRACT i ON SEPTEMBER 24, 1983, DURING ROUTINE POWER 03ERATION, THE CONTROL

ROOM OPERATOR OBSERVED THAT NO.13 OVERTEMPER4TURE DELTA TEMPERATURE INSTRUMENT SETPOINT WAS READING SLIGHTLY HIGHER THAN ALLOWED BY THE TECH SPEC. THE CHANNEL WAS DECLARED INOPERABLE AND LIMITING CONDITION 4 FOR OPERATION 3.3.1.1 ACTION 2 WAS ENTERED. THE REDUNDANT OVERTEMPERATURE DELTA TEMPERATURE CHANNELS WERE OPERABLE THROUGHOUT 4 THE OCCURRENCE. DUE TO A PROTECTION SYSTEM CHANNEL SETPOINT BEING

! FOUND OUT-OF-SPECIFICATION, THE EVENT IS REPORTABLE PER TECH SPEC 4

6.9.1.9A. INVESTIGATION REVEALED THAT THE CALIBRATION OF SUMMATOR ITC-431A AND FUNCTION GENERATOR 10M-431A HAD DRIFTED. NO PREVIOUS PROBLEMS OF THIS TYPE HAD BEEN NOTED. THE SUMMATOR, FUNCTION GENERATOR CHANNEL INDICATOR WERE RECALIBRATED; THE CHANNEL WAS SATISFACTORILY TESTED; AND THE ACTION STATEMENT WAS TERMINATED.

FORM 4 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1981 072 1 8109080304 168710 07/23/81 l

DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE i ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG

< ABSTRACT WHILE PERFORMING THE S.S.P.S. PERIODIC TEST, THE OPERATOR DISCOVERED THAT LOGIC A SWITCH IN POSITION 1 DID NOT FUNCTION PROPERLY IN MODES BLOCKS INHIBIT AND NOT INHIBITED. SWITCil POSITION 1 TESTS THE LOSS OF FLOW LOOP 1 CIRCUITRY. S.S.P.S. TRAIN B WAS DECLARED INOPERABLE.

UNIVERSAL BOARD SERIAL NO. 0156 HAD FAILED. UNIVERSAL BOARD SERIAL l NO. 0156 WAS REPLACED IN KIND BY UNIVERSAL BOARD SERIAL NO. 0220.

S.S.P.S. TRAIN B WAS TESTED SATISFACTORILY AND RETURNED TO SERVICE. l

) FORM 5 LER SCSS DATA 07-22-86 l

j DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE l

1

, -------- -,m-,,,,,,, ,--a--, nn-,.--.-,. ~w,---w_nn---,n- . , - - - - , ---n

=

j 311 1981 095 0 8110020402 168957 08/28/81 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG ABSTRACT IT WAS DISCOVERED THAT PRESSURIZER PRESSURE PROTECTION CHANNEL 1 WAS IN THE TRIPPED CONDITION, ACCOMPANIED BY A LOW PRESSURIZER PRESSURE, ONE OUT OF THREE, ANNUNCIATOR ALARM. HOWEVER, THE CHANNEL 1 INDICATOR PROVIDED PROPER INDICATION IN AGREEMENT WITH THE OTHER TWO CHANNELS.

THIS INDICATED THAT THE MALFUNCTION EXISTED IN THE PROTECTION CHANNEL ITSELF. PRESSURIZER! PRESSURE CHANNEL 1 WAS PLACED IN THE TRIPPED CONDITION WITHIN ONE' HOUR. UNIVERSAL BOARD A417 IN THE SOLID STATE PROTECTION SYSTEM, JRAIN A. WAS REPLACED. THE CHANNEL WAS TESTED SATISFACTORILY AND/ RETURNED TO SERVICE.

FORM 6 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1981 120 0 8112230295 171644 11/13/81 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPE 1ATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG ABSTRACT WHILE PERFORMING A SURVEILLANCE TEST ON THE SOLID STATE PROTECTION SYSTEM, TRAIN B THE LOGIC SYSTEM FOR THE REACTOR COOLANT LOOP FLOW FAILED THE LOGIC TEST. TRAIN B WAS DECLARED INOPERABLE. THE CAUSE OF THE LOGIC TEST FAILURE WAS THE FAILURE OF AN A303 UNIVERSAL CIRCUIT BOARD. THE FAILED A303 UNIVERSAL CIRCUIT BOARD WAS REPLACED AND TESTED SATISFACTORILY. TRAIN 8 SURVEILLANCE TEST WAS SATISFACTORILY PERFORMED.

FORM 7 LE SCSS DATA 07-22-86

- - - - r.***********************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1982 008 0 8203090658 172617 01/28/82 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE APCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

o .

SYMBOL: PEG ABSTRACT THE OPERATOR NOTICED THAT THE OVERPOWER DELTA T INDICATION FOR CHANNEL 21 HAD FAILED HIGH. THE CHANNEL WAS DECLARED INOPERABLE AND ACTION STATEMENT 3.3.1 ACTION 6 WAS ENTERED. THE CHANNEL WAS PLACED IN THE TRIPPED CONDITION WITHIN ONE HOUR. ALL TECH. SPEC. REQUIREMENTS WERE MET. FIVE CAPACITORS WERE REPLACED IN THE UNIT, AND THE UNIT WAS BENCH TESTED SATISFACTORILY. FUNCTIONAL TEST 2PD2.6,001 WAS SATISFACTORILY PERFORMED. CHANNEL 21 WAS DECLARED OPERABLE, AND ACTION STATEMENT 3.3.1 ACTION 6 WAS TERMINATED.

FORM 8 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1982 066 0 8208190011 175858 07/20/82 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS C0.

SYMBOL: PEG ABSTRACT DURING TEST FOR NO. 21 STEAM GENERATOR STEAM FLOW CHANNEL II, THE HIGH TRIP SETPOINT WAS FOUND TO EXCEED THE TECH SPECS LIMIT. THE CHANNEL WAS DECLARED INOPERABLE, AND LIMITING CONDITIONS FOR OPERATION 3.3.1 ACTION 7 AND 3.3.2 ACTION 14 WERE ENTERED. THE EVENT CONSTITUTED OPERATION IN A DEGRADED MODE IN ACCORDANCE WITH TECH SPECS 6.9.1.9.B.

SEE LERS: 82-061,82-054, 82-052,82-042, 82-018. THE CHANNEL TRIP SETPOINT WAS OUT-OF-SPECIFICATION DUE TO SEVERAL FAILED CAPACITORS IN MODULES 2FC-513 AND 2FM-513E. THE FAILURES WERE ATTRIBUTED TO NATURAL DEGENERATION OVER TIME. ALL CAPACITORS IN THE MODULES WERE REPLACED.

THE FUNCTIONAL TEST WAS SATISFACTORILY COMPLETED AND THE ACTION STATEMENT TERMINATED.

FORM 9 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1982 085 0 8209220302 176810 08/16/82 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG ABSTRACT TWO OCCURRENCES OF HALFUNCTION OF NO. 24 REACTOR COOLANT CHANNEL FLOW

O .

INDICATION WERE EXPERIENCED. THE FIRST WAS AT 1240 HOURS. THE CHANNEL WAS DECLARED IN0PERABLE AND ACTION STATEMENT 3.3.1 ACTION 7 WAS ENTERED AT 1245 HOURS. THE ISOLATION AMPLIFIER WAS REPLACED AND TESTED SATISFACTORILY. THE CHANNEL WAS DECLARED OPERABLE AND ACTION STATEMENT 3.3.1 ACTION 7 WAS TERMINATED AT 1407 HOURS. THE SECOND OCCURRENCE WAS AT 2330 HOURS. ACTION STATEMENT 3.3.1 ACTION 7 WAS ENTERED AT 2330 HOURS. THE ISOLATION AMPLIFIER WAS REPLACED AND TESTED SATISFACTORILY, AND ACTION STATEMENT 3.3.1 ACTION 7 WAS TERMINATED AT 0005 HOURS, AUGUST 17, 1982. BOTH OCCURRENCES WERE ATTRIBUTED TO FAILURE OF THE SIGNAL ISOLATION AMPLIFIERS. THIS CAUSED INACCURATE SIGNALS TO BE SENT TO THE CHANNEL FLOW METER. THE AMPLIFIERS WERE REPLACED AND TESTED SATISFACTORILY. NO COMMON CAUSE FOR SIGNAL ISOLATION AMPLIFIER FAILURE COULD BE DETERMINED.

FORM 10 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1982 125 0 8211300227 179456 10/13/82 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS C0.

SYMBOL: PEG ABSTRACT DURING THE PERFORMANCE OF THE SOLID STATE PROTECTION SYSTEM (SSPS)

FUNCTIONAL TEST, A TECHNICIAN DISCOVERED THAT THE TEST SEQUENCE FOR THE STEAMLINE ISOLATION FUNCTION ON TRAIN B FAILED TO OPERATE. DUE TO THE FAILURE, LIMITING CONDITION FOR OPERATION 3.3.2 ACTION 20 APPLIED. THE CHANNEL WAS ALREADY BYPASSED FOR PERFORMANCE OF THE SURVEILLANCE, AND WAS MAINTAINED IN THAT CONDITION UNTIL REPAIRS WERE COMPLETED. SSPS TRAIN A WAS OPERABLE THROUGHOUT THE OCCURRENCE. THE EVENT CONSTITUTED OPERATION IN A DEGRADED MODE IN ACCORDANCE WITH TECH SPEC 6.9.1.9.B. FAILURE OF THE STEAMLINE ISOLATION FUNCTION WAS DUE TO A FAILED CIRCUIT BOARD. THE FAILED BOARD WAS REPLACED, AND THE FUNCTIONAL TEST WAS SATISFACTORILY COMPLETED.

FORM 11 LER SCSS DATA 07-22-86

- - - - a***********

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1982 154 0 8301190361 180745 12/13/82 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS C0.

SYMBOL: PEG

o a ABSTRACT DURING ROUTINE PERFORMANCE OF THE CHANNE FUNCTIONAL TEST, AN INSTRUMENT TECHNICIAN OBSERVED THAT NO. 22 STEAM GENERATOR STEAM PRESSURE CHANNEL III BISTABLE SETPOINT WAS LESS THAN THE 480 PSIG REQUIRED BY THE TECH SPECS. THE CHANNEL WAS DECLARED IN0PERABLE AND LIMITING CONDITION FOR OPERATION 3.3.2 ACTION 14 WAS ENTERED. THE BISTABLES ASSOCIATED WITH THE CHANNEL WERE ALREADY TRIPPED. THE EVENT CONSTITUTED OPERATION IN A DEGRADED MODE IN ACCORDANCE WITH TECH SPEC 6.9.1.9.B. THE PROBLEM WAS ATTRIBUTED TO AN ISOLATED FAILURE OF A CAPACITOR IN COMPARATOR 2PC 526 C/D. ALL CAPACITORS IN THE MODULE WERE REPLACED, THE INSTRUMENT WAS SATISFACTORILY TESTED, AND THE ACTION STATEMENT WAS TERMINATED.

FORM 12 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 311 1983 055 0 8311180203 187121 10/04/83 DOCKET:311 SALEM 2 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: PSEG FACILITY OPERATOR: PUBLIC SERVICE ELECTRIC & GAS CO.

SYMBOL: PEG ABSTRACT ON OCTOBER 4, 1983, DURING ROUTINE POWER OPERATION, THE CONTROL ROOM OPERATOR OBSERVED THAT THE N0. 21 STEAM GENERATOR PRESSURE CHANNEL II INSTRUMENT INDICATED EXCESSIVELY DIFFERENT THAN THE REDUNDANT CHANNELS. CHANNEL II WAS DECLARED IN0PERABLE AND LIMITING CONDITION FOR OPERATION 3.3.2 ACTION 14 WAS ENTERED. THE REDUNDANT CHANNELS WERE OPERABLE THROUGHOUT THE OCCURRENCE; THE EVENT CONSTITUTED OPERATION IN A DEGRADED MODE PER TECH SPEC 6.9.1.98. INVESTIGATION PEVEALED NO APPARENT PROBLEM WITH THE TRANSMITTER. THE CALIBRATION OF THE DEVICE WAS CHECKED AND MINOR ADJUSTMENTS WERE MADE. UPON RESTORING THE CHANNEL TO SERVICE IT INDICATED WITHIN TOLERANCE OF THE REDUNDANT CHANNELS. NO FURTHER PROBLEMS WERE NOTED; THE CHANNEL WAS DECLARED OPERABLE AND THE ACTION STATEMENT WAS TERMINATED.

FORM 13 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 315 1981 043 0 8110270444 169285 09/19/81 DOCKET:315 COOK 1 TYPE:PWR REGION: 3 NSSS:WE

ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME

ABSTRACT DURING THE PERFORMANCE OF A TEST ON THE SOLID STATE PROTECTION SYSTEM (TPAIN A), A LOGIC TEST POSITION TESTED UNSATISFACTORILY. THIS AFFECTED THE CONTAINMENT PRESSURE HI-HI PORTION OF THE AUTOMATIC ACTUATION LOGIC. UNIVERSAL CARD A-210 (WESTINGHOUSE, PART NO.

60E6D21G01) WAS FOUND TO HAVE FAILED. THE LOGIC CARD WAS REPLACED, THE SURVEILLANCE TEST PROCECURE WAS PERFORMED ON TRAIN A TO VERIFY SYSTEM OPERABILITY.

FORM 14 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 315 1983 042 0 8306210142 183331 05/12/83 DOCKET:315 COOK 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING A SURVEILLANCE TEST ON THE PRESSURIZER PRESSURE PROTFCTION SET III, IT WAS FOUND THAT THE HIGH PRESSURE REACTOR TRIP BISTABLE WOULD NOT TRIP. THIS EVENT WAS NON-CONSERVATIVE WITH RESPECT TO TECH SPEC 3.3.1.1 TABLE 3.3-1 ITEM 10. THE ACTION REQUIREMENTS WERE MET. THIS IS THE FIRST OCCURRENCE OF THIS TYPE. INVESTIGATION REVEALED THAT THE HIGH PRESSURE REACTOR TRIP BISTABLE, MANUFACTURED BY FOXBOR0 COMPANY HAD FAILED. THE BISTABLE WAS REPLACED, CALIBRATED, VERIFIED TO BE OPERATING CORRECTLY AND RETURNED TO SERVICE.

FORM 15 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 315 1983 081 0 8309160212 185645 08/15/83 00CKET:315 COOK 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING A SURVEILLANCE TEST ON THE SOURCE RANGE INSTRUMENTATION, PRIOR TO HEAD LIFT, IT WAS INDICATED THAT THERE WAS A MALFUNCTION IN THE LOOP 3 OVERTEMPERATURE DELTA-T CHANNEL III CIRCUITRY. THIS EVENT OCCURRED IN A MODE WHERE THE REACTOR TRIP SYSTEM INSTRUMENTATION IS NOT APPLICABLE, BUT IS BEING REPORTED SINCE THIS EQUIPMENT FAILURE IS IN SAFETY RELATED INSTRUMENTATION. DURING AN APPLICABLE MODE THIS EVENT WOULD HAVE BEEN NONCONSERVATIVE WITH RESPECT TO TECH SPEC r

i

3.3.1.1, TABLE 3.3-1 ITEM 7. REQUIREMENTS OF THE ACTION ITEMS WERE MET. PREVIOUS OCCURRENCES INCLUDE: 050-315/78-067,81-043 AND 50-316/78L-076,82-056,83-033,038,047. UNIVERSAL LOGIC CARD A-313 (WESTINGHOUSE, PART NO. 6056D21G01) WAS FOUND TO HAVE FAILED. THE LOGIC CARD WAS REPLACED, THE SURVEILLANCE TEST WAS PERFORMED ON TRAIN "B" TO VERIFY SYSTEM OPERABILITY. NO FURTHER ACTION IS PLANNED.

FORM 16 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 315 1983 113 0 8312090073 187546 11/17/83 DOCKET:315 COOK 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT AT 0910 HOURS DURING PLANNED SURVEILLANCE TESTING, LOOP 1 OVERPOWER DELTA TEMPERATURE ROD STOP AND REACTOR TRIP BISTABLE FAILED TO TRIP.

AT 1150 HOURS, LOOP 2 OVERTEMPERATURE DELTA TEMPERATURE MODULE FAILED HIGH. THESE EVENTS COMBINED WERE LESS CONSERVATIVE THAN THE LEAST CONSERVATIVE REQUIREMENT OF TECH SPEC 3.3.1.1, TABLE 3.3-1. THE ACTION REQUIREMENTS WERE MET. THE LOOP 1 BISTABLE, OVERPOWER DELTA TEMPERATURE R00 STOP AND REACTOR TRIP WAS REPLACED, CALIBRATED, AND RETURNED TO SERVICE AT 1226 HOURS. AT THIS TIME THE UNIT WAS PLACED IN A CONDITION THAT WAS WITHIN THE REQIUREMENTS OF ACTION STATEMENT #6 0F TECH SPEC 3.3.1.1 TABLE 3.3-1. THE MODULE FOR LOOP 2 OVERTEMPERATURE DELTA TEMPERATURE WAS ALSO REPLACED, CALIBRATED AND RETURNED TO SERVICE AT 1901 HOURS.

FORM 17 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1981 074 0 8201200637 172092 12/15/81 DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING THE PERFORMANCE OF A SURVEILLANCE TEST ON THE SOLID STATE PROTECTION SYSTEM (TRAIN B), THREE LOGIC TEST POSITIONS TESTED UNSATISFACTORILY. THIS AFFECTED THE CONTAINMENT PRESSURE HI-HI PORTION OF THE AUTOMATIC ACTUATION LOGIC. SIMILAR OCCURRENCES INCLUDE: 050-315/78-067,81-043 AND 050-316/78-076-080,79-005.

, , [ s y .

h 3(3 l y ;

V

'l UNIVERSAL CARD A-209 (WESTINGHOUSE, JART NU. 6056D21G01) WAS FOUND TO HAVE FAILED. THE LOGIC CARD WAR REPLACED,'THE SURVEILLANCE TEST PROCEDURE WAS PF.RFORMED ON TRAIh "B" TO VERIFY SYSTEM OPEARSILITY. NO FURTHER ACTION IS PLANNED. ,.

FORM 18 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER' NSIC EVENT DATE s 316 1982 044 0 8206170226 174075 05/12/82

- - - - +********+**********

DOCKET:316 COOK 2 TYPE:PWR

. REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER:'AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT

, THE LOOP 4 REACTOR COOLANT LOSS OF FLOW BISTABLE (FB-445A) SETPOINT WAS FOUND TO HAVE EXCEEDED LCO SETP0ff!T. THIS EVENT WAS NONCONSERVATIVE IN RESPECT TO TECH SPEC TABLE 2.2-1 ITEM 12 AND 3.3.1.1 TABLE 3.3-1 ITEM 12. THECAUSEOFTHE. BISTABLE (FB-445A)

EXCEEDING THE LC0 SETPOINT WAS INSTRUMENT DRIFf. THE BISTABLE, MANUFACTURED BY FOXBOR0, MODEL 630-AC-OHAAF WAS REPLACED, CALIBRATED,

>AND THE LOOP WAS RETURNED TO SERVICE.

FORM 19 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 19EL. 064 1 8210040209 176590 07/24/82

\ DOCKET:316 COOK 2 TYPE:PWR

! REGION: 3 NSSS:WE

~ ARCHITECTURAL ENGINEER: AEPS .

, FACILITY OPERATOR: INDIANA & MICHIGAN E!.ECTRIC C0.

SYMBOL:-IME 1 <

ABSTRACT LOOP 3 OVERTEMPERATURE DELTA T TRIP SETPOINT WAS FOUND TO HAVE DRIFTED

, HIGH. THIS EVENT WAS NON-CONSERVATIVE WITH RESPEr.T TO TECH SPEC ,

3 M.1.1 TABLE 3.3-1 ITEM 7. INVE!rIGATION FOUND THAT THE LOOP 3 OVLRTEMPERATURE DELTA T SETPCINT LGERATOR MODULE HAD DRIFTED OUT OF SPECIFICATION. THE MODULE, MANUFAf,TURED BY FOXBOR0, MODEL NO.

66RC-OLA SPECIAL WAS RECALIBRATED MD WITHIN A FEU HOURS WAS AGAIN ll0T '

OF SPECIFICATION. A NEW MODULE WA'3 INSTALLED, CALIBRATED, VERIFIED TO BE OPERATING CORRECTLY ANC.THE LOOP WAS RETURNED TO SERVICE.

\

FORM 20 LER SCSS DATA 07-22-86 r \ ~

\

4

o. _ . _ _ _ _ _ _ _ _ . _ _ _ - . _ - __. _ __-

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 033 0 8304110759 182357 03/08/83 DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS .

FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING THE PERFORMANCE OF A SURVEILLANCE TEST ON THE SOLID STATE PROTECTION SYSTEM (TRAIN B) A LOGIC TEST POSITION TESTED UNSATISFACTORILY. THIS AFFECTED THE POWER RANGE HIGH FLUX RATE TRIP PORTION OF THE AUTOMATIC LOGIC, TECH SPEC TABLE 3.3-1 ITEM 22. THE REQUIREMENTS OF ACTION ITEM 1 WERE MET. PREVIOUS OCCURRENCES INCLUDE:

050-315/78-067,81-043 AND 050-316/78-076,82-056. UNIVERSAL CARD B-212 (WESTINGHOUSE, PART NO. 6056D21G01) WAS FOUND TO HAVE FAILED.

THE LOGIC CARD WAS REPLACED, THE SURVEILLANCE TEST PROCEDURE WAS PERFORMED ON TRAIN 'B' TO VERIFY SYSTEM OPERABILITY. NO FURTHER ACTION IS PLANNED.

FORM 21 LER SCSS DATA 07-22-86

- - - h*w*************************************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 038 0 8305020177 182546 04/05/83 DOCKET:316 COOK 2 TYPE:PWR- .

REGI0t: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS .

FAGjlITYOPERATOR: INDIAt!A & MICHIGAN ELECTRIC C0.

SYMBOL: IME as ABSTRACT DURING A SURVEILLANCE TEST ON THE SOLID STATE PROTECTION SYSTEM (TRAIN A), FOUR LOGIC TEST POSITIONS TESTED UNSATISFACTORILY. THESE AFFECTED THE LOSS OF REACTOR COOLANT FLOW IN LOOP 2 AND P-8 PERMISSIVES. THIS EVENT WAS NON-CONSERVATIVE WITH RESPECT TO TECH SPECS TABLE 3.3-l' ITEM 22. AFTER THE ONE HOUR TIME LIMIT, (ACTION 1

ITEM 1), A POWER REDUCTION WAS STARTED WHILE TROUBLESHOOTING CONTINUED. PREVIOUS OCCURRENCES INCLUDE: 050-315/78-067,81-043 AND

. 50-316/78-076,82-056,83-033. UNIVERSAL CARD A-304 (WESTINGHOUSE, PART N0. 6056021G01) WAS FOUND TO HAVE FAILED. THE LOGIC CARD WAS REPLACED, THE SURVEILLANCE TEST PROCEDURE WAS PERFORMED ON TRAIN "A" TO VERIFY SYSTEM OPERABILITY. NO FURTHER ACTION IS PLANNED.

FORM 22 LER SCSS DATA . 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 047 0 8306060271 182928 05/03/83

. o 00CKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING A SURVEILLANCE TEST ON THE SOLID STATE PROTECTION SYSTEM (TRAIN B), A LOGIC TEST POSITION TESTED UNSATISFACTORILY. THIS AFFECTED THE POWER RANGE HIGH FLUX RATE TRIP PORTION OF THE AUTOMATIC LOGIC, TECH SPEC TABLE 3.3-1 ITEM 22. THE REQUIREMENTS OF THE ACTION ITEM WERE MET. PREVIOUS OCCURRENCES INCLUDE: 050-315/78-067,81-043 AND 50-316/78-076,82-056, 83-033, 038. UNIVERSAL CARD A-212 (WESTINGHOUSE, PART N0. 6056D21G01) WAS FOUND TO HAVE FAILED. THE LOGIC CARD WAS REPLACED, AND THE SURVEILLANCE TEST WAS PERFORMED ON TRAIN "B" TO VERIFY SYSTEM OPERABILITY. NO FURTHER ACTION IS PLANNED.

FORM 23 LER SC3S DATA 07-22-86

- - - - +*****************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 090 0 8310140140 185943 09/20/83

- - - - e******

DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE

. ARCHITECTURAL ENGINEER: AEPS FACILITY.0PERATOR: INDIANA & MICHIGAN ELECTRIC CO. ,

SYMBOL: IME ,

ABSTRACT ,

~

DURING A SURVEILLANCE TEST ON THE SOLID STATE PROTECTION SYSTEM (TRAIN

, A), TWO LOGIC POSITIONS TESTED UNSATISFACTORILY. THIS AFFECTED THE REACTOR COOLANT PUMP UNDERVOLTAGE POSITION OF THE AUTOMATIC LOGIC,

-- TECH SPEC TABLE 3.3-1 ITEM 22.. AFTER THE ONE HOUR TIME LIMIT, (ACTION ITEM #1), PREPARATIONS TO BE IM H0T STANDBY WITHIN SIX HOURS WERE INITIATED. PREVIOUS OCCURRENCES INCLUDE: 050-315/83-081,81-043.

78-067 AND 50-316/83-033, 038, 047,82-056, 78-076. UNIVERSAL CARD A-311 (WESTINGHOUSE, PART NO. 6056D21G01) WAS FOUND TO BE DEFECTIVE.

THE LOGIC CARD WAS REPLACED. THE SURVEILLANCE TEST WAS PERFORMED ON TRAIN "A" TO VERIFY SYSTEM OPERABILITY. H0 FURTHER ACTION IS PLANNED.

FORM 24 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 123 0 8401240191 188419 12/20/83 DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE

, o l l

ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN FLECTRIC CO.

SYMBOL: IME ABSTRACT DURING A SURVEILLANCE TEST ON STEAM GENERATOR 1 PROTECTION CHANNEL SET I, IT WAS FOUND THAT THE OUTPUT FROM BISTABLE 2FB-512B (STEAM FLOW GREATER THAN REFERENCE STEAM FLOW) FLUCTUATED AT BOTH THE TRIP AND RESET POINTS. THE BISTABLE WAS REPLACED. THIS EVENT WAS NON-CONSERVATIVE WITH RESPECT TO TECH SPEC 3.3.1.1 TABLE 3.3-1 ITEM

15. THE ACTION REQUIREMENTS WERE MET. THE REPLACEMENT BISTABLE WAS CALIBRATED, VERIFIED TO BE OPERATING CORRECTLY AND RETURNED TO SERVICE.

FORM 25 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 124 0 8401310050 188420 12/24/83 DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING SURVEILLANCE TEST ON POWER RANGE DRAWER, A CONTROL POWER FUSE BLEW RENDERING THE CHANNEL INOPERABLE. THIS IS NONCONSERVATIVE WITH RESPECT TO TECH SPEC 3.3.1.1. THE FUSE WAS REPLACED WITHIN THE ALLOWABLE TIME LIMITS. OPERATORS WERE USING A FLUKE TO OBTAIN READINGS. C&I PERSONNEL COULD NOT FIND ANY PROBLEM WITH THE DIGITAL V0LTMETER USED. THE POWER RANGE DRA'AER WAS CALIBRATED'AND NO PROBLEMS WERE ENC 0UNTERED. THE REASON FOR THE CONTROL FUSE FAILURE COULD NOT BE DETEF. MINED AND THE PROBLEM DID NOT RECUR. NO FURTHER ACTIONS PLANNES.

FORM 26 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1983 3.25 0 8401300257 188421 12/29/83 DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT DURING THE PERFORMANCE OF THE LOOP 4 POWER RANGE NUCLEAR

) O INSTRUMENTATION CALIBRATION TEST, THE OUTPUT FROM THE FLUX PENALTY MODULE (2NY-441) WAS FOUND TO BE DRIFTING IN THE CONSERVATIVE DIRECTION. THE MODULE WAS REPLACED. THIS EVENT WAS NON-CONSERVATIVE WITH RESPECT TO TECH SPEC 3.3.1.1 TABLE 3.3-1. THE ACTION REQUIREMENTS WERE MET. THE REPLACEMENT FLUX PENALTY MODULE WAS CALIBRATED, VERIFIED TO BE OPERATING CORRECTLY AND RETURNED TO SERVICE.

FORM 27 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 316 1984 030 0 8412130378 192290 11/19/84 DOCKET:316 COOK 2 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: AEPS FACILITY OPERATOR: INDIANA & MICHIGAN ELECTRIC CO.

SYMBOL: IME ABSTRACT POWER LEVEL - 096%. ON 11-19-84, UNIT 2 WAS OPERATING AT 96% REACTOR THERMAL POWER WITH THE LOOP 3 SG FEEDWATER LEVEL BISTABLES FOR RPS CHANNEL _III TRIPPED. THE BISTABLES WERE IN A TRIPPED CONDITION DUE TO THE IN0PERABILITY OF THE LOOP 3 SG FEEDWATER LEVEL INDICATION. AT APPROX 0356 HRS, A SPURIOUS ACTUATION OF THE RPS FEEDWATER/ STEAM FLOW MISMATCH BISTABLE FOR THE LOOP 3 SG OCCURRED. THIS TRIPPED BISTABLE, IN COMBINATION WITH THE EXISTING TRIPPED CONDITION OF THE SG LEVEL BISTABLE, ACTIVATED THE RPS LOGIC AND PRODUCED A REACTOR TRIP. THE BISTABLE AND AS5OCIATED FEEDWATER/ STEAM FLOW TRANSMITTERS WERE

_. SUBSEQUENTLY TESTED BUT NO REASON FOR THE SPURIOUS ACTUATION WAS FOUND. CONSEQUENTLY, NO CORRECTIVE ACTIONS ARE PLANNED.

FORM 28 LER SCSS DATA . 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 327 1981 007 0 8102100557 163891 01/16/81 DOCKET:327 SEQUOYAH 1 TYPE:PWR f

REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: TVAX FACILITY OPERATOR: TENNESSEE VALLEY AUTHORITY SYMBOL: TVA ABSTRACT THE LEVEL CHANNEL FOR STEAM GENERATOR #3 (L-3-97) WAS DECLARED INOPERABLE WHEN THE TRIP BISTABLE (1-LS-3-97F) FAILED TO DEENERGIZE DURING THE PERFORMANCE OF THE SCHEDULED FUNCTIONAL TEST. THE LEVEL CHANNEL FAILED BECAUSE THE BISTABLE WOULD NOT DEENERGIZE. THE MODULE WAS REPLACED, CALIBRATED, AND RETURNED TO SERVICE.

l l

FORM 29 LER SCSS DATA 07-22-86

e.4****************************************************************

DCfAET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 327 1981 058 1 8107310508 167738 05/22/81 DOCKET:327 SEQUOYAH 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: TVAX FACILITY OPERATOR: TENNESSEE VALLEY AUTHORITY SYMBOL: TVA ABSTRACT SURVEILLANCE TESTING ON S/G LEVEL CHANNEL 1-L-3-38 IDENTIFIED THAT THE LO-L0 LEVEL BISTABLE SETP0INTS FOR REACTOR TRIP AND SAFETY FEATURES ACTUATION WERE .1432 VOLTS (ALLOWABLE LIMIT IS /= .180 V0LTS). THE BISTABLE WAS TRIPPED. NO PREVIOUS OCCURRENCES. THE EVENT WAS CAUSED BY A COMPONENT FAILURE. THE BISTABLE WAS REPLACED, CALIBRATION WAS COMPLETED, AND THE INSTRUMENT LOOP WAS RETURNED TO SERVICE.

FORM 30 LER SCSS DATA . 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 327 1981 121 0 8111030544 169934 09/25/81 DOCKET:327 SEQUOYAH 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: TVAX FACILITY OPERATOR: TENNESSEE VALLEY AUTHORITY' .

SYMBOL:'TVA ABSTRACT THE FLUX IMBALANCE PENALTY FUNCTION GENERATOR'(XM-92-5005K) WAS DECLARED IN0PERABLE DUE TO BEING OUT OF CALIBRATION. THE 70XBOR0 MANUFACTURED FLUX IMBALANCE PENALTY FUNCTION GENERATOR ('l-XM-92-5005K)

OUT OF TOLERANCE WAS DUE TO MODULE DRIFT. THE MODULE WAS RECALIBRATED AND RETURNED TO SERVICE. .

FORM 31 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 327 1983 150 0 8312010240 187733 10/25/83 DOCKET:327 SEQUOYAH 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: TVAX FACILITY OPERATOR: TENNESSEE VALLEY AUTHORITY SYMBOL: TVA

ABSTRACT WITH UNIT 1 IN MODE 1 (100% REACTOR POWER) AT 0915 ON 10/25/83, ONE CHANNEL OF REACTOR COOLANT PUMP UNDERFREQUENCY TRIP WAS DECLARED INOPERABLE DUE TO FAILURE TO MEET SURVEILLANCE TESTING. THIS EVENT REQUIRED ENTRY INTO ACTION STATEMENT '6' 0F LCO 3.3.1.1. DURING FUNCTIONAL TESTING 0F TRAIN 'A' RCP UNDERFREQUENCY TRIP, A UNIVERSAL LOGIC CIRCUIT BOARD (A215) LOCATED IN THE SSPS FAILED MOST PROBABLY DUE TO A BAD CONNECTION. THE UNIVERSAL CIRCUIT CARD WAS REPLACED AND THE CHANNEL SATISFACTORILY TESTED AT 0821 ON 10/25/83. NO FURTHER ACTION IS PLANNED.

FORM 32 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 334 1981 057 0 8107240108 167614 06/10/81 DOCKET:334 BEAVER VALLEY 1 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: DUQUESNE LIGHT C0.

SYMBOL: DUQ ABSTRACT A GROUNDED CONDITION CAUSING A PROCESS CABINET ALARM WAS BEING COPRECTED WHEN A FAILED POWER SUPPLY FOR THE RATE MODULE OF THE LOOP C MAIN STEAM BREAK PROTECTION DURING C00LDOWN CHANNEL 3 WAS DISCOVERED.

THE CHANNEL WAS DECLARED IN0PERABLE. A BAD TRANSISTOR WAS FOUND ON THE POWER SUPPLY AND RATE CARD. THE ENTIRE CARD WAS REPLACED AND THE CHANNEL RETURNED TO SERVICE.

' FORM 33 LER SCSS DATA ~07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE l

l 334 1981 074 0 8109110402 168812 08/03/81 DOCKET:334 BEAVER VALLEY 1 TYPE:PWR REGION: 1 NSSS:WE l ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: DUQUESNE LIGHT C0.

SYMBOL: DUQ ABSTRACT ESF INSTRUMENTATION PS-MS-475A (BISTABLE FOR LOOP A, CHANNEL III, LOW

('

l STEAM LINE PRESSURE) ALARMED AND RESET NUMEROUS TIMES AND WAS DECLARED INOPERABLE UPON VERIFICATION THAT STEAM LINE PRESSURE WAS NORMAL. A l

CIRCUIT BOARD CAPACITOR FAILED DAMAGING AN ADJACENT CIRCUIT BOARD.

BOTH CIRCUIT BOARDS WERE REPLACED.

l

O O FORM 34 LER SCSS DATA 07-22-86

- - - - j DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE l 334 1982 037 0 8210150208 179097 09/21/82 DOCKET:334 ~ BEAVER VALLEY 1 TYPE:PWR REGION: 1 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: DUQUESNE LIGHT CO.

SYMBOL: DUQ ABSTRACT STEAM GENERATOR (S/G) 1A LEVEL PROTECTION CHANNEL III SIGNAL COMPARATOR (LC-FW476A) WAS DECLARED IN0PERABLE DUE TO FAULTY ACTUATION AND TRIP SETP0 INT DRIFT FOUND DURING SURVEILLANCE TESTING. THE PROTECTION CHANNEL BISTABLE WAS MANUALLY TRIPPED PRIOR TO THE TEST AND LEFT IN THIS CONDITION AS REQUIRED BY TECH SPEC. THE INCIDENT RESULTED FROM THE FAILURE OF THE HAGAN COMPARATOR. THE CAUSE IS UNKNOWN AT THIS TIME. ON 9/21/82 AFTER REPLACING THE COMPARATOR, THE CHANNEL WAS RETURNED TO SERVICE. DUE TO THE PAST RELIABILITY OF THE HAGAN CONTROL INSTRUMENTATION, THIS FAILURE IS CONSIDERED AN ISOLATED INCIDENT.

FORM 35 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 334 1983 029 0 8311030126 186972 09/18/83 D0'CKET:334 BEAVER VALLEY 1 TYPE:PWR.

REGION: I NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: DUQUESNE LIGHT C0.

SYMBOL: DUQ ABSTRACT .

ON 9/18/83, WHILE IN HOT STANDBY, WITH RCS PRESSURE AT 1400 PSI, MAIN STEAM PRESSURE TRANSMITTER (PT-MS-484) WAS DECLARED INOPERABLE AFTER IT WAS DISCOVERED TO BE READING ZER0 WHILE THE OTHER TWO REDUNDANT CHANNELS INDICATED 400 PSIG. THE IN0PERABLE CHANNEL WAS PLACED IN THE TRIPPED CONDITIONS WITHIN ONE HOUR AS PER THE ACTION STATEMENT REQUIREMENTS OF TECH SPEC 3.3.2.1. UPON INVESTIGATION, THE PRESSURE TRANSMITTER'S IN0PERABILITY WAS ATTRIBUTED TO ITS OSCILLATOR / AMPLIFIER BEING OUT OF CALIBRATION. (PT-MS-484) WAS RETURNED TO SERVICE AT 2024 HOURS WITH RCS PRESSURE AT 2230 PSI AFTER THE OSCILLATOR / AMPLIFIER WAS RECALIBRATED.

FORM 36 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE

338 1981 043 0 8106230473 166506 05/23/81 DOCKET:338 NORTH ANNA 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: VEP ABSTRACT C STEAM GENERATOR FEEDWATER FLOW CHANNEL IV DROPPED TO APPR0XIMATELY 85% OF ITS PREVIOUSLY INDICATED VALUE. THE AFFECTED CHANNEL WAS PLACED IN TRIP WITHIN ONE HOUR AS REQUIRED BY TECH SPEC. THE CAUSE OF THIS EVENT WAS INSTRUMENT DRIFT IN THE SQUARE ROOT EXTRACTOR (MULTIPLIER - DIVIDER) CARD. THE CHANNEL WAS PLACED IN TRIP, THE CARD RECALIBRATED, AND TESTED SATISFACTORILY.

FORM 37 LER SCSS DATA 07-22-86

+*******************************************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 338 1981 054 0 8108030118 167556 07/03/81 DOCKET:338 NORTH ANNA 1 TYPE:PWR RECION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT .'

~~

IT WAS DISCOVERED THAT THE FLOW INDICATION FOR THE 3A AUXILIARY FEEDWATER PUMP FAILED TO INDICATE FLOW DURING THE MONTHLY SURVEILLANCE TEST. THE PUMP DISCHARGE FLOW WAS VERIFIED BY A DECREASE IN MAIN FEEDWATER FLOW WITH STABLE STEAM GENERATOR LEVEL. AN INSTRUMENT CARD ..

IN THE POWER SUPPLY FOR FT-FW-100C FAILED CAUSING THE INPUT TO THE FLOW INDICATOR TO DROP TO 0 V0LTS. CONSEQUENTLY, THE INDICATOR DID NOT SHOW A FLOW. THE POWER SUPPLY CARD WAS REPLACED AND THE FLOW INDICATION WAS VERIFIED TO FUNCTION CORRECTLY.

FORM 38 LER'SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 338 1981 066 1 8112010416 170255 08/04/81 00CKET:338 NORTH ANNA 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP l l

l l

ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: VEP ABSTRACT CHANNEL III FEEDWATER FLOW INDICATION FOR LOOP 2 FAILED HIGH. THIS CONDITION WOULD HAVE PREVENTED THE CHANNEL FROM GENERATING A REACTOR TRIP SIGNAL ON A STEAM FLOW / FEED FLOW MISMATCH (FS FW) COINCIDENT WITH LOW STEAM GENERATOR LEVEL. THE AFFECTED STEAM FLOW FEED FLOW REACTOR TRIP BISTABLE WAS PLACED IN THAT TRIPPED CONDITION IN 1 HOUR BY PLACING THE FEED FLOW CHANNEL IN TEST. THE FEED FLOW CHANNEL FAILED HIGH DUE TO FAILURE OF THE LOOP MULTIPLIER / DIVIDER / SQUARE ROOT CARD. THE DEFECTIVE NMD CARD WAS REPLACED AND THE CHANNEL WAS SATISFACTORILY RECALIBRATED AND RETURNED TO SERVICE.

FORM 46 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1981 047 0 8107310431 167640 06/09/81 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: VEP ABSTRACT DURING THE REACTOR PROTECTION AND ESF LOGIC TEST, THE POSITIONING 0F -

THE LOGIC TEST SWITCH IN THE SSPS TRAIN B RESULTED IN A LOGIC-FAULT LIGHT ON THE TEST PANEL INDICATING A FAILURE IN THE LOGIC CIRCUIT.

INVESTIGATION REVEALED A FAILED UNIVERSAL CARD IN THE TRAIN. THE FAULTED CARD WAS REPLACED, THE CHANNEL CALIBRATED AND RETURNED TO SERVICE. -

9

~

FORM 47 LER SCSS DATA 07-22-86'

- ******************************************4*************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1981 048 0 8107310218 167558 06/13/81 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT INTERMEDIATE RANGE NUCLEAR INSTRUMENTATION CHANNEL N-35 FAILED ON JUNE 13, 1981 AND AGAIN ON JUNE 14, 1981. FAILURE OF A HIGH VOLTAGE POWER SUPPLY AND A LOSS OF DETECTOR V0LTAGE BISTABLE MODULE CAUSED THE

o O FAILURES OF NUCLEAR INSTRUMENTATION CHANNEL N-35. THE FAILED COMPONENTS WERE REPLACED AND NUCLEAR INSTRUMENTATION CHANNEL N-35 WAS RETURNED TO SERVICE.

FORM 48 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1982 064 0 8211040375 178862 10/12/82 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT INTERMEDIATE RANGE CHANNEL N-35 FAILED LOW. SINCE REACTOR POWER WAS GREATER THAN 5% N0 FURTHER ACTIONS WERE REQUIRED UNTIL SHUTDOWN. THIS EVENT IS CONTRARY TO TECH SPEC 3.3.1 AND REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9.B. THE CAUSE FOR INTERMEDIATE RANGE CHANNEL N-35 FAILING LOW IS THE RESULT OF A FAILED LOG CURRENT AMPLIFIER MODULE. THE MODULE WAS REPLACED, TESTED SATISFACTORILY AND RESTORED TO SERVICE.

FORM 49 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1982 067 1 8305050276 183237 10/17/82 DOCKET:339 NORTH ANNA 2 TYPE:PWR -

REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY,0PERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP..

-ABSTRACT . .

ON OCTOBER 17,-1982,~WHILE IN MODE 1, THE 'C' STEAM GENERATOR NARROW

RANGE CHANNEL 2 (LI-2495) INDICATED HIGH AND WAS DECLARED IN0PERABLE.

THE AFFECTED CHANNEL WAS PLACED~IN TRIP WITHIN 1 HOUR AS REQUIRED BY

ACTION 7 0F TECH SPEC 3.3.1.1. THIS EVENT IS REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9. THE INSTRUMENT DRIFT WAS DUE TO THE FAILURE OF THE TRANSMITTER AMPLIFIER CARD. THE CARD WAS REPLACED, THE LOOP RECALIBRATED AND THE CHANNEL RETURNED TO SERVICE.

FORM 50 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1983 012 0 8302280491 182041 01/14/83 .

- - - - =********************************************.************

00CKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: VEP ABSTRACT ON JANUARY 14, 1983 WHILE OPERATING AT 30% POWER, THE DELTA T INDICATOR (TI-412A) ASSOCIATED WITH DELTA T PROTECTION FOR LOOP A, WAS READING APPR0XIMATELY 6% HIGH. THE AFFECTED DELTA T/TAVE CHANNEL WAS PLACED IN THE TRIP CONDITION WITHIN ONE HOUR. THIS EVENT IS CONTRARY TO TECH SPEC 3.3.1.1 AND REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9.B.

THE HIGHER THAN NORMAL READING ON THE DELTA T INDICATOR WAS CAUSED BY A FAILED RELAY CARD ON THE OUTPUT OF TE-24128 (T HOT LOOP A). THE CARD WAS REPLACED, CALIBRATED AND RETURNED TO SERVICE.

FORM 51 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1983 015 0 8303210465 182375 02/02/83 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT ON FEBRUARY 2, 1983, WITH UNIT 2 AT 100% POWER, THE RESET VALUE FOR THE ENGINEERING SAFETY FEATURES (ESF) P-12 INTERLOCK WAS FOUND OUT OF LIMIT HIGH ON ONE OF THREE CHANNELS. THIS INTERLOCK IS ASSOCIATED

'.- WITH THE DELTA T/T-AVG PROTECTION CHANNEL III. THE OTHER TWO DELTA T/TAVG PROTECTION CHANNELS REMAINED AVAILABLE. THIS EVENT IS CONTRARY TO TECH SPEC 3.3.2.1 AND REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9.A.

THE ENGINEERING SAFETY FEATURES P-12 INTERLOCK RESET VALUE ON ONE OF THREE CHANNELS WAS HIGH BECAUSE OF AN ERRATIC AND SENSITIVIE POTENTIOMETER ON COMPARATOR CARD TC-432E. ATTEMPTS TO ADJUST THE DEVICE WERE NOT SUCCESSFUL. THE AFFECTED COMPARATOR CARD WAS .

REPLACED, TESTED AND RESTORED TO SERVICE.

~

FORM 52 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1983 025 0 8305030283 182258 03/31/83

- - - - 4*********************

DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

, e-SYMBOL: VEP ABSTRACT ON MARCH 31, 1983 WITH THE UNIT IN MODE 1, IT WAS DISCOVERED THAT THE STEAM GENERATOR "B" WATER LEVEL, CHANNEL I, HIGH-HIGH TURBINE TRIP /FEEDWATER ISOLATION SETPOINT WAS GREATER TRAN THE ALLOWABLE VALUE OF 76 PERCENT REQUIRED BY TECH SPEC 3.3.2.1(5.A.). THE AFFECTED CHANNEL WAS PLACED IN THE TRIP CONDITION AND THE TWO REDUNDANT WATER LEVEL CHANNELS FOR "B" STEAM GENERATOR WERE OPERABLE. THIS EVENT IS REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9.A. THE CAUSE OF THIS EVENT WAS A MALFUNCTIONING COMPARATOR CARD. THE HIGH-HIGH LEVEL COMPARATOR CARD WAS REPLACED, CALIBRATED TO THE CORRECT SETPOINT AND THE CHANNEL WAS RETURNED TO SERVICE. THIS EVENT HAS NO GENERIC IMPLICATIONS.

FORM 53 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1983 039 0 8306090314 183267 05/13/83 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT ON MAY 13, 1983, THE UNDERVOLTAGE (UV) OUTPUT BOARD OF TRAIN "B" SOLID STATE PROTECTION SYSTEM WAS FOUND FAILED DURING A CHANNEL FUNCTIONAL

, TEST. THIS DISABLED THE AUTOMATIC REACTOR TRIP FUNCTION OF TRAIN B.

THE REDUNDANT TRAIN OF SSPS WAS CAPABLE OF PROVIDING AN AUTOMATIC REACTOR TRIP IF REQUIRED AND MANUAL REACTOR TRIP CAPABILITIES WERE AVAILABLE. THIS EVENT IS CONTRARY TO TECH SPEC 3.3.1.1 AND PEPORTABLE PURSUANT T0 TECH SPEC 6.9.1.9.B. THE CAUSE OF THE EVENT WAS A FAILED

. UV OUTPUT BOARD. THE BOARD WAS SUBSEQUENTLY REPLACED AND THE CHANNEL FUNCTIONALLY TESTED SATISFACTORILY.

FORM 54 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1983 079 0 8401040365 187954 12/11/83 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT ON DECEMBER 11, 1983, WITH UNIT 2 IN MODE 1, PRESSURIZER LEVEL

, e INDICATION, CHANNEL II LI-2460 FAILED LOW. THE REMAINING TWO CHANNELS WERE STILL OPERABLE AND THE FAILED CHANNEL WAS PLACED IN TRIP WITHIN ONE HOUR TO ASSURE CONSERVATIVE COINCIDENCE OF REACTOR PROTECTION ACTUATION SIGNALS. THIS EVENT IS CONTRARY TO TECH SPEC 3.3.1.1 AND REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9.B. THE EVENT WAS CAUSED BY THE FAILURE OF THE POWER SUPPLY CARD FOR LI-2460. THE CARD WAS REPLACED AND THE CHANNEL CALIBRATED SATISFACTORILY. NO FURTHER CORRECTIVE ACTION IS REQUIRED. THERE ARE N0 GENERIC IMPLICATIONS OF THIS EVENT.

FORM SS LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 344 1983 021 0 8402030368 188697 12/20/83 DOCKET:344 TROJAN TYPE:PWR REGION: 5 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: PORTLAND GENERAL ELECTRIC C0.

SYMBOL: PGC ABSTRACT ON DECEMBER 20, 1983 THE CONTROL RODS FAILED TO RESPOND TO A DEMAND FOR R00 MOTION WHILE THE RODS WERE IN MANUAL CONTROL. THE PROBLEM WAS DETERMINED TO BE A FAILED MASTER PULSER CARD IN THE R0D CONTROL SYSTEM.' THIS PLACED THE PLANT IN A DEGRADED MODE PERMITTED BY A LIMITING CONDITION FOR OPERATION (TECH SPEC 3.1.3.1). THE RODS WERE CAPABLE OF BEING TRIPPED BOTH MANUALLY AND AUTOMATICALLY DURING THE TWO AND ONE-HALF HOUR PERIOD THAT THE R0D CONTROL SYSTEM WAS IN0PERABLE. THE CAUSE OF THIS OCCURRENCE WAS A FAILED MASTER PULSER CARD IN THE R0D CONTROL SYSTEM. THE CARD WAS REPLACED WITHIN TWO AND ONE-HALF HOURS. THE RODS WERE CAPABLE OF BEING TRIPPED'IF NECES$ARY DURING THIS PERIOD. ,

i 1

07-22-86 l FORM "56 LER SCSS DATA DOCKET YEAR LER NUMBER REVISION DCS NUF"ER NSIC EVENT DATE 344 1985 001 0 8503140a19 195238 11/09/84 l

D0CKET:344 TROJAN TYPE:PWR REGION: 5 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: PORTLAND GENERAL ELECTRIC Cr SYMBOL: PGC ABSTRACT POWER LEVEL - 100%. DURING A REVIEW 0F THE SEISMIC QUALIFICATION FOR THE AUXILIARY FEEDWATER CONTROL VALVES (CV-3004A1,81,C1,DI,A2,B2,C2, AND D2) IN NOVEMBER 1984, IT WAS DISCOVERED THAT THE VALVE CONTROL l

l

l UNITS MAY NOT BE FULLY SEISMICALLY QUALIFIED. DOCUMENTATION DOES NOT )

EXIST TO SHOW THAT THE VALVE POSITION CONTROL ELEMENTS WERE !

FUNCTIONALLY TESTED AFTER EXCITATION DURING SEISMIC QUALIFICATION TESTING. ALTHOUGH THE CONTROLLER DESIGN IS SUCH THAT THE MOST LIKELY PROBABILITY OF A SEISMIC INDUCED FAILURE WCULD BE FOR THE VALVE TO FAIL APPR0XIMATELY 50% OPEN, IT MUST BE CONSERVATIVELY POSTULATED THAT ALL EIGHT VALVES COULD FAIL CLOSED. THE LACK 0F COMPLETE SEISMIC QUALIFICATION WAS THEN DETERMINED TO BE REPORTABLE ON 01/31/85 UNDER 10CFR50.73(A)(2)(V) SINCE IT IS A CONDITION WHICH COULD HAVE AFFECTED THE CAPABILITY OF A SAFETY SYSTEM FROM REMOVING RESIDUAL HEAT OR MITIGATING THE CONSEQUENCES OF AN ACCIDENT. NO SEISMIC EVENTS OF SIGNIFICANT MAGNITUDE (GREAT EN0 UGH TO ACTIVATE THE SEISMIC EVENT RECORDER) HAVE OCCURED AT TROJAN. EVEN IF A SEISMIC EVENT DID OCCUR, AND EVERY ONE OF THE EIGHT FEEDWATER FLOW CONTROL VALVES FAILED CLOSED, THE VALVES CAN BE MANUALLY REPOSITIONED AS REQUIRED. IN ADDITION, PLANT PROCEDURES ARE IN PLACE TO REC 0VER FROM A TOTAL LOSS OF SECONDARY HEAT SINK. CORRECTIVE ACTION WILL BE TO REPLACE THE NON-SEISMICALLY QUALIFIED COMPONENTS WITH SEISMICALLY QUALIFIED EQUIVALENTS.

FORM 57 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE

*348

* * *1981 n. 0 8105080289

- - - - ,,','*************** 166032 * * * *04/05/81 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT WHILE PERFORMING A CHANNEL CHECK SURVEILLANCE TEST, THE OVERTEMPERATURE - DELTA TEMPERATURE INSTRUMENTATION LOOP ASSOCIATED WITH TI 412C WAS DECLARED INOPERABLE WHEN ITS TRIP SETP0 INT FAILED TO MEET THE TOLERANCE REQUIREMENTS OF THE TEST. THIS OCCURRENCE IS ATTRIBUTABLE TO A DEFECTIVE NLL CARD AND AN OUT OF ADJUSTMENT GAIN POTENTIOMETER. THE DEFECTIVE CARD WAS REPLACED, THE POTENTIOMETER ADJUSTED, AND THE CHANNEL WAS RETURNED TO SERVICE.

FORM 58 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1981 018 0 8105190361 166219 04/13/81 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT WHILE PERFORMING INTERMEDIATE RANGE FUNCTIONAL CHECK CHANNEL N-35/36, INTERMEDIATE RANGE CHANNEL N-35 WAS DECLARED IN0PERABLE WHEN ITS CURRENT METER READING FAILED TO MEET THE TEST REQUIREMENTS. INCIDENT WAS CAUSED BY A DEFECTIVE LOG CURRENT AMPLIFIER CARD. THE CARD WAS REPLACED, THE APPLICABLE PORTIONS OF FNP-1-STP-41.2 SATISFACTORILY PERFORMED AND N-35 WAS RETURNED TO SERVICE ON 4-14-81.

FORM 59 LER SCSS DATA 07-22-86

- - - - +***********************************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1981 025 0 8106030240 166290 05/02/81 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE

, ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT THE PRESSURIZER PRESSURE INDICATOR, 457C, WAS REMOVED FROM SERVICE DUE TO A FAULTY LEAD / LAG CARD. AS A RESULT, PI457C INSTRUMENTATION LOOP WAS DECLARED IN0PERABLE. THE OCCURRENCE WAS CAUSED BY A FAULTY LEAD / LAG CARD. THE CARD WAS REPLACED AND FOLLOWING CALIBRATION THE INSTRUMENTATION LOOP, PI457C, WAS RETURNED TO SERVICE.

FORM 60 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER HSIC EVENT DATE ~

348 1981 060 0 8109220525 168865 08/12/81 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT THE A LOOP TAVG. PROTECTION CHANNEL (412D) WAS DECLARED INOPERABLE WHEN THE ASSOCIATED TEMPERATURE INDICATOR (TI-412D) FAILED LOW. THIS EVENT WAS CAUSED BY THE FAILURE OF A LEAD-LAG CARD. THE CARD WAS REPLACED AND FOLLOWING CALIBRATION, PROTECTION CHANNEL 412D WAS RETURNED TO SERVICE.

FORM 61 LER SCSS DATA 07-22-86 l

l

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1982 037 0 8208200173 175896 07/18/82

- - - - +**************************************************

DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT ON 7-18-82, THE B OVERPOWER - DELTA TEMPERATURE INSTRUMENTATION LOOP ASSOCIATED WITH TE-422 WAS DECLARED IN0PERABLE WHEN BISTABLES TS/4228-1 AND TS/422B-2 TRIPPED FOR NO APPARENT REASON. ON 7-20-82, THE INSTRUMENTATION LOOP WAS DECLARED IN0PERABLE WHEN ERRATIC READINGS ON THE TE-422 INDICATOR WERE NOTED. TECH SPEC 3.3.1, IN PART, REQUIRES THIS INSTRUMENTATION CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS DUE TO A FAULTY AMPLIFIER CARD. INVESTIGATION FOLLOWING THE FIRST OCCURRENCE DID NOT REVEAL THE CAUSE OF THE EVENT. FOLLOWING SUCCESSFUL COMPLETION OF FNP-1-STP 201.19 (REACTOR COOLANT SYSTEM TE422B AND TE4220 LOOP CALIBRATION AND FUNCTIONAL TEST), THE INSTRUMENTATION LOOP WAS DECLARED OPERABLE. FOLLOWING THE SECOND OCCURRENCE, THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-201.19 THE INSTRUMENTATION LOOP WAS DECLARED OPERABLE.

FORM 62 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1982 054 0 8210260463 177739 09/20/82

~ DOCKET:348 FARLEY 1 TYPE:PWR .

REGION: 2 NSSS:WE ,

ARCHITECTURAL ENGINEER: BESS -

, FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT THE INSTRUMENTATION LOOP ASSOCIATED WITH FEEDWATER FLOW TRANSMITTER FT-497 WAS DECLARED IN0PERABLE WHEN THE FLOW TRANSMITTER FAILED LOW.

TECH SPEC 3.3.1, IN PART, REQUIRES THIS INSTRUMENTATION LOOP TO BE j OPERABLE. THIS EVENT WAS DUE TO THE FAILURE OF AN NLP PRINTED' CIRCUIT CARD IN THE 7300 PROCESS CONTROL SYSTEM. THE CARD WAS REPLACED AND FOLLOWING CALIBRATION AND FUNCTIONAL TESTS,' INSTRUMENTATION LOOP FT-497 WAS RETURNED TO SERVICE.

l FORM 63 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 001 0 8302070620 181076 01/02/85 l

DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0445 AND 1545 ON 1/2/83, REACTOR COOLANT SYSTEM SUBC00 LING MONITOR CHANNEL 2 WAS DECLARED INOPERABLE FOLLOWING THE RECEIPT OF MAIN CONTROL BOARD ALARMS. TECH SPEC 3.3.3.8, IN PART, REQUIRES THE RCS SUBC00 LING MONITOR CHANNEL TO BE OPERABLE. TECH SPEC 3.3.3.8 ACTION STATEMENT REQUIREMENTS WERE MET. THESE EVENTS WERE ATTRIBUTED TO A FAULTY PRINTED CIRCUIT CARD. THE CONNECTOR PINS ON THE CARD WERE CLEANED AND IT WAS RETURNED TO SERVICE AT 1520 ON 1/2/83. FOLLOWING THE SECOND OCCURRENCE. THE CARD WAS REPLACED, AND UPON SATISFACTORY PERFORMANCE OF FNP-1-STP-201.32A (CORE COOLING MONITOR CHANNEL CHECK),

RCS SUBC00 LING MONITOR CHANNEL 2 WAS DECLARED OPERABLE AT 0007 ON 1/3/83.

FORM 64' LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 012 0 8305050158 182565 04/02/83 DOCKET:348 FARLEY 1 TYPE:PWP REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS '

FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ,,

ABSTRACT c AT 1813 ON 4/2/83, 0VERTEMPERATURE-DELTA TEMPERATURE LOOP 2 CHANNEL II WAS DECLARED IN0PERABLE WHEN BISTABLES 422C(1) AND 422C(2) TRIPPED.

TECH SPECS 3.3.1, IN PART, REQUIRES THIS INSTRUMENTATION CHANNEL TO BE OPERABLE. TECH 3PEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET.

THIS EVENT WAS DUE TO A POWER SUPPLY FAILURE OF A COMPARATOR CARD.

FOLLOWING REPLACEMENT OF THE CARD AND CALIBRATION OF THE INSTRUMENTATION LOOP, OVERTEMPERATURE-DELTA TEMPERATURE LOOP 2 CHANNEL II WAS DECLARED OPERABLE AT 0340 ON 4/3'83.

FORM 65 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 013 0 8400000000 188626 03/29/83 DOCKET:348 FARLEY 1 TYPE:PWP REGION: 2 NSSS:WE

ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 2140 ON 3/29/83, THE INSTRUMENTATION LOOP ASSOCIATED WITH STEAM GENERATOR 1A LEVEL TRANSMITTER LT-475 WAS DECLARED IN0PERABLE DUE TO A LOW INDICATION. TECH SPEC 3.3.1 IN PART, REQUIRES THE LT-475 INSTRUMENTATION LOOP TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS DUE TO THE FAILURE OF A POWER SUPPLY CARD. THE CARD WAS REPLACED AND FOLLOWING THE PERFORMANCE OF FNP-1-STP-213.26 (STEAM GENERATOR 1A 01C22LT0475 LOOP CALIBRATION AND FUNCTIONAL TEST), THE LT-475 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0132 ON 3/30/83.

FORM 66 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 020 0 8305250581 182952 04/20/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 2137 ON 4/20/83, THE "B" OVERTEMPERATURE DELTA TEMPERATURE INSTRUMENTATION LOOP ASSOCIATED WITH TI422C WAS DECLARED IN0PERABLE DUE TO ITS READING BEING OUT OF TOLERANCE WITH THE OTHER CHANNELS.

TECH SPEC 3.3.1, IN PART, REQUIRES THE INSTRUMENTATION LOOP TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET.

THIS EVENT WAS DUE TO DRIFT OF THE SUMMING AMPLIFIER CARD. THE CARD WAS ADJUSTED AND FOLLOWING PERFORMANCE OF FNP-1-STP-201.19 (REACTOR COOLANT SYSTEM TE-422B AND TE-422D LOOP CALIBRATION AND FUNCTIONAL TEST), "B" OVERTEMPERATURE DELTA TEMPERATURE LOOP WAS DECLARED .

OPERABLE AT 0317 ON 4/21/83.

i i FORM 67 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC. EVENT DATE 348 1983 028 0 8400000000 188627 06/01/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC

a a ABSTRACT AT 0415 ON 6/1/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH IC STEAM GENERATOR FLOW TRANSMITTER FT-495 WAS DECLARED INOPERABLE DUE TO ITS ERR 0NE0US INDICATION. TECH SPEC SECTIONS 3.3.1 AND 3.3.2, IN PART, REQUIRE THIS INSTRUMENTATION CHANNEL TO BE OPERABLE. TECH SPEC ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS DUE TO DRIFT OF THE SQUARE ROOT EXTRACTION CARD. THE CARD WAS ADJUSTED AND FOLLOWING THE COMPLETION OF FNP-1-STP-213.24 (STEAM GENERATOR IC 01C22FT0495 LOOP CALIBRATION AND FUNCTIONAL TEST), FT-495 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0845 GN 6/1/83.

FORM 68 LER SCSS DATA 07-22-86

- - - - kt**********************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 029 0 8400000000 188628 06/01/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0416 ON 6/1/83 THE OVERPOWER DELTA TEMPERATURE CHANNEL I WAS DECLARED IN0PERABLE UPON RECEIPT OF AN OVERPOWER DELTA T SINGLE INPUT ALERT WHEN AN ACTUAL OVERPOWER CONDITION DID NOT EXIST. TECH SPEC 3.3.1, IN PART, REQUIRES THIS INSTRUMENTATION TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS DUE TO THE FAILURE OF THE OVERPOWER DELTA TEMPERATURE BISTABLE CARD. THE CARD WAS REPLACED AND FOLLOWING PERFORMANCE OF FNP-1-STP-201.18A (REACTOR C0OLANT SYSTEM TE412A AND TE412D LOOP CALIBRATION AND FUNCTIONAL TEST), THE OVERPOWER DELTA TEMPERATURE CHANNEL I WAS DECLARED OPERABLE AT 0528 ON 6/1/83.

FORM 69 LER SCSS DATA 07-22 ********************************************************************

NSIC EVENT DATE DOCKET YEAR LER NUMBER REVISION DCS NUMBER 348 1983 031 0 8307150109 184486 06/10/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 1130 ON 6/10/83 DURING THE PERFORMANCE OF FNP-1-STP-33.0A (SOLID STATE PROTECTION SYSTEM OPERABILITY TEST), THE LOOP 3 STEAM LINE DIFFERENTIAL PRESSURE INSTRUMENTATION CHANNEL IN THE SOLID STATE

l l

PROTECTION SYSTEM WAS DECLARED INOPERABLE UPON RECEIPT OF A FAILURE LIGHT. TECH SPEC 3.3.2, IN PART, REQUIRES THIS INSTRUMENTATION CHANNEL TO BE OPERABLE. TECH SPEC 3.3.2 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS ATTRIBUTED TO THE FAILURE OF THE UNIVERSAL CARD. FOLLOWING REPLACEMENT OF THE CARD AND SATISFACTORY PERFORMANCE OF FNP-1-STP-33.0A, THE LOOP 3 STEAM LINE DIFFERENTIAL PRESSURE INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1210 ON 6/10/83.

FORM 70 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 036 0 8307290170 184389 06/29/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 0130 ON 6/29/83, 1B STEAM GENERATOR WIDE RANGE LEVEL INDICATOR ON THE REMOTE SHUTDOWN PANEL WAS DECLARED INOPERABLE DUE TO ERRATIC READINGS. TECH SPEC 3.3.3.5, IN PART, REQUIRES THE STEAM GENERATOR LEVEL INSTRUMENTATION CHANNEL TO BE OPERABLE. TECH SPEC 3.3.3.5 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS DUE TO THE FAILURE OF A PRINTED CIRCUIT CARD. THE CARD WAS REPLACED AND F0LLOWING THE SATISFACTORY PERFORMANCE OF FNP-1-STP-30.0 (HOT SHUTDOWN PANEL INSTRUMENTATION CHANNEL CHECK), THE IB STEAM GENERATOR WIDE RANGE LEVEL INDICATOR WAS DECLARED OPERABLE AT 0741 ON 6/29/83.

FORM 71 LER SCSS DATA 07.-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 040 0 8308260166 185250 07/20/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ,

ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0948 ON 7/20/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH STEAM GENERATOR IB LEVEL TRANSMITTER LT-486 WAS DECLARED IN0PERABLE DUE TO A LOSS OF THE POWER SUPPLY TO FR0 CESS CABINET #3. TECH EDEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY A FAILURE OF THE POWER SUPPLY ON THE SIGNAL COMPARATOR CARD. THE CARD WAS REPLACED AND FOLLOWING SUCCESSFUL PERFORMANCE OF FNP-1-STP-213.6A

%. i (STEAM GENERATOR 1B Q1C22LT486 LOOP CALIBRATION AND FUNCTIONAL TEST).

THE LT-486 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1415 ON 7/20/83.

FORM 72 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 043 0 8309130421 185897 08/08/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0815 ON 8/8/83, DURING THE PERFORMANCE OF FNP-1-STP-215.5 (MAIN FEEDWATER N1C22FT0496 LOOP CALIBRATION AND FUNCTIONAL TEST), THE INSTRUMENTATION CHANNEL ASSOCIATED WITH BISTABLE FB498A (LOOP 3 STEAM FLOW-FEED FLOW MISMATCH) WAS DECLARED INOPERABLE WHEN THE BISTABLE FAILED TO TRIP. TECH SPEC 3.3.1, IN PART, REQUIRE THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET.

THIS EVENT WAS CAUSED BY FAILURE OF THE BISTABLE COMPARATOR CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-215.5, THE FB498A INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1100 ON 8/8/83.

FORM 73 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 045 0 8309130424 185898 08/10/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0821 ON 8/10/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FLOW INDICATOR FI-484 (LOOP 2 STEAM FLOW) WAS DECLARED IN0PERABLE WHEN FI-484 FAILED LOW. TECH SPEC 3.3.1 AND TECH SPEC 3.3.2, IN PART, REQUIRE THIS CHANNEL TO BE OPERABLE. TECH SPEC ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY FAILURE OF A POWER SUFPLY CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PFRFORMANCE OF FNP-1-STP-213.21 (STEAM GENERATOR 1B Q1C22FT0484 LOOP CALIBRATION AND FUNCTIONAL TEST), THE FI-484 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1322 ON 8/10/83.

h a FORM. 74 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 054 0 8309210451 185668 08/17/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 1800 ON 8/17/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH REACTOR COOLANT SYSTEM FLOW TRANSMITTER FT-434 WAS DECLARED IN0PERABLE DUE TO ERRATIC INDICATION. TECH SPEC 2.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY THE FAILURE OF AN NLP CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-201.13 (REACTOR COOLANT SYSTEM Q1821FT0434 LOOP CALIBRATION AND FUNCTIONAL TEST), THE FT-434 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0830 ON 8/22/83.

FORM 75 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 055 0 8309260369 185899 08/20/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE APCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT _

AT 1705 ON 8/20/83 AND AGAIN AT 1815 ON 8/29/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FEEDWATER FLOW INDICATOR FI-437 WAS DECLARED IN0PERABLE WHEN IT DRIFTED HIGH. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. NO CAUSE COULD BE DETERMINED F02;THE FIRST EVENT. FNP-1-215.4 (MAIN FEEDWATER FT-487 LOOP CALIBRATF?N AND FUNCTIONAL TEST) WAS PERFORMED SATISFACTORILY AND THE FI-As?

INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0852 ON.8/h?/83. .

FOLLOWINGTHESECONDEVENT,ITWASDETERMINEDTHATTHETRANSK1TTER AMPLIFIER CARD WAS INTERMITTENTLY DRIFTING. THE CARD WAS REPt\ ACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-215.4, THE FI-187 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1236 ON 8/30/8 FORM 76 LER SCSS DATA 07-22-L'5

o .

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 060 0 8310120356 186044 09/12/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 2109 ON 9/12/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FLOW TRANSMITTER FT-415 (RCS LOOP 1) WAS DECLARED IN0PERABLE DUE TO ITS ERRATIC INDICATION. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY FAILURE OF AN NLP CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-201.8 (REACTOR COOLANT SYSTEM FT-415 LOOP CALIBRATION AND FUNCTIONAL TEST),

THE FT-415 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1445 ON 9/16/83.

FORM 77 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 061 0 8310120363 186045 09/13/83 00CKET:348 FAPLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0015 ON 9/13/83, DURING PERFORMANCE OF FNP-1-STP-201.1 (PRESSURIZER LEVEL Q1B31LT0459 CALIBRATION AND FUNCTIONAL TEST), THE INSTRUMENTATION CHANNEL ASSOCIATED WITH LEVEL TRANSMITTER LT-459 (PRESSURIZER LEVEL - NARROW RANGE) WAS DECLARED IN0PERABLE WHEN IT WAS DETERMINED TO BE OUT OF TOLERANCE. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERASLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY FAILURE OF AN NLP CARD. THE CARD WAS REPLACED AllD FOLLOWING SATISFACTORY COMPLETION OF FNP-1-STP-201.1, THE LT J59 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0430 ON 9/13/83. .

FORM 78 LER SCSS DATA 07-22-86

- - - - r************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 072 0 8311220363 187013 10/13/83 l

I

_ _- . _ _ _ - -. a- . _ . __ __ ____ ___.

DOCKET:348 FARLEY 1 TYPE:PWR REGION: '2 NSSS:WE

. ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 2045 ON 10/13/83, 1330 ON 10/14/83 AND 1345 ON 10/17/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FLOW TRANSMITTER FT-495 (LOOP 3 STEAM FLOW) WAS DECLARED INOPERABLE DUE TO ERR 0NEDUS INDICATION.

TECH SPEC 3.3.1 AND 3.3.2 IN PART, REQUIRE THIS CHANNEL TO BE OPERABLE.' TECH SPEC ACTION STATEMENT REQUIREMENT 3 WERE MET. NO CAUSE FOR THESE EVENTS COULD BE DETERMINED. FOLLOWING THE FIRST EVENT, THE MSIV'S WERE CLOSED (FT-495 -INSTRUMENTATION CHANNEL NO LONGER REQUIRED TO BE OPERABLE)~AND THE LCO. CLEARED AT 2105 ON 10/13/83. WHEN MSIV'S WERE OPENED DURING SUBSEQUENT STARTUP (1330 ON 10/14/83), A SECOND LC0 WAS DECLARED. FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-213.18 (HIGH STEAM LINE FLOW STEAM LINE ISOLATION & P-13 FUNCTIONAL TEST FB-475A, FB-485A, FB-495A AND PB-447E) AND VERIFICATION OF PROPER INDICATION DURING PERFORMANCE OF FNP-1-STP-1.0 (OPERATIONS DAILY AND SHIFT SURVEILLANCE REQUIREMENTS), THE FT-495 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0545 ON 10/15/83. FOLLOWING THE THIRD EVENT, THE CHANNEL WAS CHECKED AND RETURNED TO SERVICE. AFTER VERIFICATION OF PROPER INDICATION DURING PERFORMANCE OF FNP-1-STP-1.0, THE FT-495 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0933 ON 10/18/83.

FORM 79 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 074 0 8311280380 187564 10/18/83 DOCKET:348 FARLEY I TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 0825 ON 10/18/83, THE INSTRUMENTATION CHANNELS ASSOCIATED WITH RTDS TE-412A (LOOP 1 HOT LEG) AND TE-412D (LOOP 1 COLD LEG) WERE DECLARED INOPERABLE DUE TO ERR 0NE0US INDICATION. TECH SPEC 3.3.1, IN PART, REQUIRES THESE CHANNELS TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY FAILURE OF AN AMPLIFIER CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-201.18 (REACTOR COOLANT SYSTEM TE-412A AND TE-412D LOOP CALIBRATION AND FUNCTIONAL TEST), THE TE-412A AND TE-4120 INSTRUMENTATION CHANNELS WERE DECLARED OPERABLE AT 1635 ON 10/18/83.

[90"****!h***************bk!*khh!*!$h**********************0~!!*!!

e O DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 348 1983 095 0 8402030456 188698 12/26/83 DOCKET:348 FARLEY 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 2220 ON 12/26/83, THE LOOP 1 OVERTEMPERATURE DELTA TEMPERATURE INSTRUMENTATION CHANNEL WAS DECLARED IN0PERABLE WHEN TEMPERATURE INDICATOR TI-412C (LOOP 1 OVERTEMPERATURE SETPOINT) FAILED LOW. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY FAILURE OF A SUMMING AMP (NSA) CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-1-STP-201.18 (REACTOR COOLANT SYSTEM TE-412 LOOP CALIBRATION AND FUNCTIONAL TEST), THE LOOP 1 OVERTEMPERATURE DELTA TEMPERATURE INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0400 ON 12/27/83.

FORM 81 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1981 041 1 8112010429 170244 09/30/81 DOCKET:364 FARLEY 2 TYPE:PWR .

REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT ON 9/30/81 AND ON 10/29/81, WHILE PERFORMING PRESSURIZER LEVEL LT-461 LOOP CALIBRATION AND FUNCTIONAL TEST, THE INSTRUMENTATION LOOP ASSOCIATED WITH LT-461 WAS DECLARED IN0PERABLE WHEN THE COMPARATOR.

CARD AND POWER SUPPLY CARD FAILED TO MEET THE V0LTAGE SETPOINT '

CRITERIA. THE FIRST EVENT WAS ATTRIBUTED TO COMPONENT DRIFT ON THE COMPARATOR CARD. THE CARD WAS RECALIBRATED AND FOLLOWING THE SUCCESSFUL COMPLETION OF THE TEST, THE LT-461 INSTRUMENTATION LOOP WAS RETURNED TO SERVICE ON 9/30/81. FURTHER INVESTIGATION REVEALED A FAULTY LOOP POWER SUPPLY CARD WHICH WAS REPLACED ON 10/3/81. THEs SECOND EVENT WAS AGAIN CAUSED BY COMPONENT DRIFT ON THE COMPARATOR CARD. THIS TIME, THE COMPARATOR CARD WAS REPLACED AND FOLLOWING THE SUCCESSFUL COMPLETION OF THE TEST, THE LT-461 INSTRUMENTATION LOOP WAS RETURNED TO SERVICE ON 10/29/81.

FORM 82 LER SCSS DATA 07-22-86

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1981 047 0 8112240133 171791 11/09/81 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT THE 28 STEAM GENERATOR PRESSURE INSTRUMENTATION LOOP TO THE HOT SHUTDOWN PANEL WAS DECLARED IN0PERABLE WHEN THE ASSOCIATED INDICATOR (PI3371B) FAILED LOW. THIS EVENT WAS CAUSED BY THE FAILURE OF THE LOOP POWER SUPPLY CARD. THE CARD WAS REPLACED AND FOLLOWING A SATISFACTORY FUNCTIONAL TEST, THE INSTRUMENTATION LOOP WAS RETURNED TO SERVICE.

FORM 83 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1981 051 0 8201120303 171831 12/05/81

- - - - +***********************

DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT WHILE PERFORMING FNP-2-STP-201.20B (RCS TE-4328 LOOP CALIBRATION AND FUNCTIONAL TEST) THE OVERTEMPERATURE - DELTA TEMPERATURE INSTRUMENTATION LOOP ASSOCIATED WITH TE-432 WAS DECLARED INOPERABLE WHEN BISTABLE TB-432C-1 WAS FOUND OUT OF TECH SPEC TOLERANCE. TECH ,

SPEC 3.3.1, IN PART, REQUIRES THE INSTRUMENTATION LOOP'T0 BE OPEPABLE.

THIS EVENT WAS DUE TO A COMPARATOR CARD BEING OUT OF CALIBRATION.

THE CARD WAS RECALIBRATED AND FOLLOWING THE SUCCESSFUL COMPLETION OF FNP-2-STP-201.20B, INSTRUMENTATION LOOP TE-432 WAS RETURNED TO SERVICE.

FORM 84 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1982 035 0 8209200171 176757 08/14/82 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC -

ABSTRACT REACTOR COOLANT SYSTEM SUBC00 LING MONITOR CHANNEL 1 WAS DECLARED Ih0PERABLE DUE TO INVALID MAIN CONTROL BOARD METER READINGS. TECH SPEC 3.3.3.8, IN PART, REQUIRES THIS RCS SUBC00 LING MONITOR CHANNEL TO BE OPERABLE. THIS EVENT WAS CAUSED BY A FAULTY PRINTED CIRCUIT CARD.

THE PRINTED CIRCUIT CARD WAS REPLACED AND FOLLOWING CALIBRATION RCS SUBC00 LING MONITOR CHANNEL 1 WAS DECLARED OPERABLE.

FORM 85 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1982 037 0 8210010147 176845 09/05/82 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT WHILE PERFORMING FNP-2-STP-33.0 (SOLID STATE PROTECTION SYSTEM RELIABILITY TEST), THE A TRAIN AUTO TRIP LOGIC FOR HIGH STEAM GENERATOR DIFFERENTIAL PRESSURE WAS DECLARED IN0PERABLE WPEN A FAILED CARD INDICATION WAS RECEIVED BY THE AUTOMATIC TESTER. TEC;l SPEC 3.3.2, IN PART, REQUIRES THIS TRIP LOGIC TO BE OPERABLE. THIS EVENT WAS DUE TO THE FAILURE OF THE UNIVERSAL CARD (A317). THE CARD WAS REPLACED AND FOLLOWING THE SUCCESSFUL COMOLETION OF FNP-2-STP-33.0, THE A TRAIN AUTO TRIP LOGIC FOR HIGH STEAN GENERATOR DIFFERENTIAL PRESSURE WAS RETURNED TO SERVICE.

FORM 86 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 004 0 8400000000 188794 02/01/83

, DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS .

FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ,

ABSTRACT AT 0912 ON 2/1/83, THE INSTRUMENTATION CHANNEL ASSOCAITED WITH TAVG i INDICATOR TI-412D WAS DECLARED INOPERABLE WHEN TI-412D FAILED HIGH.

TECH SPEC 3.3.1, IN PART, REQUIRES THIS INSTRUMENTATION CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET.

l THIS EVENT WAS DUE TO THE FAILURE OF A TRANSISTOR ON A LEAD LAG CARD.

l l

~

l

(.1 s ,- ;'

s' s o 3 L s

THE TRANSISTOR WAS REPLACED AND FOLLOWING THE SATISFACTORY PERFORMANCE OF FNP-2-STP,,201, 18 (REACTOR COOLANT SYSTEM TE-412B AND TE-412D LOOP CALIBP,ATION AhD FUNCTIONAL TEST). THE TI-412D a '

INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1050 ON 2/1/83.

s i FORM 87 LER SCSS DATA 07-22-86 .

- - - - n *************************************** -

o DOCKET YEAR LEPsNUMBER REVISION DCS rd.MBER NSIC EVENT DATE 364 1983 '005 0 8400000000 188795 02/05/83

- - - - f=****************************

\

00CKET:364 FARLEY 2 TYPE:PWR s

REGION: E2 NSSS:WE .

ARCHITECTURAL ENGINEER: ,RESS FACILITY OPERATOR: /.LACANA POWER CO. '

? SYMBOL: APC s ABSTRACT AT 2340 ON 2/5/83, DURING'THE PERFORMANCE OF FNP-2-STP-213.1 (STEAM GENERATOR 2A LT-474 LOOP CALIBRA7:0N AND FUNCTIONAL TEST), THE INSTRUMENTATION LOOP ASSOCIATED hlTH LEVEL TRANSMITTER LT-474 WAS '

g DECI.ARED INOPERABLE DUE TO ITS TPIP SETPOINT BEING OUT OF TOLERANCE.

TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL"T0 BE OPEPABLE. < TECH-SPEC'3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS DUE TO THE FAILURE OF A COMPARATOR CARD. THE CARD WAS REPLACED AND FOLLOWING THE PERFORMANCE OF FNP-2-STP-213.1, THE LT-474 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0200 ON s 2/6/83.

t 07-22-86 FORM 88 LER SCSS DATA DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC s EVENT DATE 364 1983 008 0 8400000000' 188798 02/09/83

- - - - k***************************************************

i lYPE:PWR \

00CKET:364 FARLEY 2 REGION: 2 NSSS:WE !'

ARCHITECTURAL ENGINEER: BESS '

s

[

i* FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC

, i <

ABSTRACT ' '

! AT 1730 ON 2/9/83, DURING THE PERFORMANCE OF FNP-2-STP-33.0 { SOLID STATE; PROTECTION SYSTEM'IRAIN B OPERABILITY TEST), "B" TRAIN POWER s

RANGE'HIGH SETP0 INT TRIP LOGIC WAS DECLARED IN0PERABLE WHEN A FAILED CARD INDICATION WAS RECEIVED BY THE AUTOMATIC TESTER. TECH SPEC i 3.3.1, IN PART, REQUIRES THIS TRIP LOGIC TO BE OPERABLE. . TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT MS CAUSED '

B7 THE FAILURE OF UNIVERSAL CARD A404. THE CARD WAS REPLACE 0 AND

. ' FOLLOWING THE SATISFACTORY- PERFORMANCE OF FNP-2-STP-33.0,, THE "B" TRAIM POWER RANGE HIGH SETP0 INT TRIF LOGIC WAS DECLARED OPERABLE;AT '

' l~ s 1800 ON 2/9/83. s i s a '

'1 l , s i >

}

l l L\ s.

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FORM 89 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 009 0 8303210468 182143 02/17/83 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 1310 ON 2/17/83, PRESSURIZER POWER OPERATED RELIEF VALVE PORV 4448 WAS DECLARED IN0PERABLE WHEN PRESSURIZER PRESSURE CONTROLLER CARD 444A FAILED LOW. TECH SPEC 3.4.5, IN PART, REQUIRES PORY 4448 TO BE OPERABLE. TECH SPEC 3.4.5 ACTION STATEMENT REQUIREMENTS WERE MET.

THIS EVENT WAS DUE TO THE FAILURE OF THE DRIVER CARD. FOLLOWING THE CALIBRATION AND INSTALLATION OF A NEW CARD, PORV 444B WAS DECLAFED OPERABLE AT 1623 ON 2/17/63. ,

FORM 90 LER SCSS LATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE' 364 1983 016 0 8304220203 182320 03/18/83 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 2300 ON 3/18/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH SERVICE WATER DISCHARGE FLOW TRANSMITTER FT-571 WAS DECLARED INOPERABLE WHEN THE MAIN CONTROL BOARD RECORDER INDICATED LOW. TECH SPEC 3.3.3.10, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.3.10 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY WATER DAMAGE TO THE FLOW TRANSMITTER WHICH IS LOCATED IN A SERVICE WATER VALVE BOX. FOLLOWING THE REPAIR OF THE TRANSMITTER, THE FT-571 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1640 GN 3/19/83. AN ENGINEERING REVIEW WILL BE MADE TO DETERMINE MEANS FOR IMPROVING RELIABILITY.

FORM 91 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS HUMBER NSIC EVENT DATE 364 1983 019 0 8305100139 182266 04/04/83 I

l

A .

(- :

DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 2205 ON 4/4/83 AND 0617 ON 4/19/83, THE ROD CONTROL SYSTEM WAS DECLARED INOPERABLE UPON RECEIPT OF A FAILURE ALARM ON POWER CABINET 2AC. TECH SPEC 3.1.3.1, IN PART, REQUIRES ALL CONTROL RODS TO BE OPERABLE.- TECH SPEC 3.1:3.1 ACTION STATEMENT REQUIREMENTS WERE MET.

A SIMILAR OCCURRENCE WAS REPORTED IN LER 83-013/03L-0. NO IMMEDIATE CAUSE FOR THE FIRST FAILURE INDICATION COULD BE DETERMINED. WHEN THE ALARM WAS RESET, THE FAILURE LIGHTS AND ALARM CLEARED. FOLLOWING SATISFACTORY PERFORMANCE OF FNP-2-STP-5.0 (FULL LENGTH CONTROL R0D OPERABILITY TEST) THE R0D CONTROL SYSTEM WAS DECLARED OPERABLE AT 0000 ON 4/5/83. FURTHER TROUBLE-SHOOTING WAS PLANNED TO IDENTIFY AND CORRECT THE ROOT CAUSE OF THE EVENT. FOLLOWING THE SECOND OCCURRENCE, PHASE CONTROL CARD B2 WAS REPLACED. THE ALARM WAS RESET AND FOLLOWING THE SATISFACTORY PERFORMANCE OF FNP-2-STP-5.0, THE R0D CONTROL SYSTEM WAS DECLARED OPERABLE AT 0842 ON 4/19/83.

FORM 92 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 020 0 8400000000 188702 04/20/83

- - - - +**********************************

DOCKET:36'4 FARLEY 2' TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0440 ON 4/20/83 DURING THE PERFORMANCE OF FNP-2-STP-213.11 (SG 2A 02N11PT0475, SG 2B 02N11PT0485, SG 2C 02N11PT0495 LOOP CALIBRATION AND FUNCTIONAL TEST), STEAM GENERATOR PRESSURE TRANSMITTFR PT485 INSTRUMENTATION LOOP WAS DECLARED INOPERABLE DUE T0 ihE OUT OF SPEC CONDITION OF THE BISTABLE TRIP SETPOINT. TECH SPEC 3.3.2, IN PART, REQUIRES THE PT485 INSTRUMENTATION LOOP TO BE OPERABLE. TECH SPEC 3.3.2 ACTION STATEMENT REQUIREMENTS WERE MET. THIS' EVENT WAS DUE TO SETP0 INT DRIFT OF A PRINTED CIRCUIT CARD. THE CARD WAS REPLACED AND FOLLOWING THE SATISFACTORY PERFORMANCE OF FNP-2-STP-213.11.PT 485 INSTRUMENTATION LOOP WAS DECLARED OPERABLE AT 1250 ON 4/20/83.

FORM 93 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 031 0 8309090575 185467 08/06/83

o ,

DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 0810 ON 8/6/83, DURING THE PERFORMANCE OF FNP-2-STP-213.1 (STEAM GENERATOR 2A LT-474 LOOP CALIBRATION AND FUNCTIONAL TEST Q2C22LT0474),

THE INSTRUMENTATION CHANNEL ASSOCIATED WITH LEVEL TRANSMITTER LT-474 WAS DECLARED INOPERABLE DUE TO ITS TRIP SETPOINT BEING OUT OF TOLERANCE. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET.

THIS EVENT WAS CAUSED BY SETPOINT DRIFT OF A COMPARATOR CARD. THE CARD WAS ADJUSTED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-2-STP-213.1, THE LT-474 INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0950 ON 8/6/83.

FORM 94 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 038 0 8400000000 188803 09/08/83 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 1000 ON 9/8/83, DURING PERFORMANCE OF FNP-2-STP-33.0 (SOLID STATE PROTECTION SYSTEM TRAIN A(B) OPERABILITY TEST), THE B TRAIN SSPS 2A STEAM GENERATOR LO-LO LEVEL INSTRUMENTATION CHANNEL WAS DECLARED INOPERABLE UPON RECEIPT OF A FAILURE LIGHT. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY FAILURE OF THE UNIVERSAL CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY COMPLETION OF FNP-2-STP-33.0, THE 2A STEAM GENERATOR LO-LO LEVEL INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 1037 ON 9/8/83.

FORM 95 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 054 1 8312200479 188876 10/28/83 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER C0.

SYMBOL: APC ABSTRACT AT 2112 ON 10/28/83 AND 1630 ON 11/3/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FLOW TRANSMITTER FT-496 (LOOP 3 FEED FLOW) WAS DECLARED IN0PERABLE DUE TO ERR 0NEOUS INDICATION. TECH SPEC 3.3.1 AND 3.3.2, IN PART, REQUIRE THIS CHANNEL TO BE OPERABLE. TECH SPEC ACTION STATEMENT REQUIREMENTS WERE MET. NO CAUSE FOR THESE EVENTS COULD BE DETERMINED. FOLLOWING SATISFACTORY PERFORMANCE OF FNP-2-STP-215.5 (MAIN FEEDWATER FT-496 LOOP CALIBRATION AND FUNCTIONAL TEST N2C22FT04c6) AND VERIFICATION OF PROPER INDICATION DURING PERFORMANCE OF FNP-2-STP-1.0 (OPERATIONS DAILY AND SHIFT SURVEILLANCE REQUIREMENTS), THE FT-496 INSTRUMENTATION' CHANNEL WAS DECLARED OPERABLE AT 0150 ON 10/29/83 AND 0500 ON 11/4/83.

FORM 95 LER SCSS DATA 07-22-86

- - - - u**********************************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 064 0 8401240275 188449 12/14/83

- - - - u***********************************************************

DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 1325 ON 12/14/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FLOW INDICATOR FI-474 (LOOP 1, STEAM FLOW) WAS DECLARED INOPERABLE DUE TO ITS DRIFTING OUT OF TOLERANCE. TECH SPECS 3.3.1 AND 3.3.2, IN PART, REQUIRE THIS CHANNEL TO BE OPERABLE. TECH SPEC ACTION STATEMENT REQUIREMENTS WERE MET. NO CAUSE FOR THIS EVENT COULD BE DETERMINED.

FOLLOWING SATISFACTORY COMPLETION OF FNP-2-STP-213.19 (STEAM GEN. 2A FT-474 LOOP CALIBRATION AND FUNCTIONAL TEST 02C22FT0474) AND VERIFICATION OF PROPER INDICATION, TliE FI-474 INSTRUMENTATION CHANNEL HAD RETURNED TO NORMAL AND WAS DFCLARED OPLR.'ILE AT 0158 ON 12/15/83.

FORM 97 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 364 1983 065 0 8401300303 188450 12/27/83 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT

m g AT 0840 ON 12/27/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH TEMPERATURE INDICATOR TI-432 (RCS LOOP 3 T-AVE) WAS DECLARED IN0PERABLE DUE TO ERR 0NE0US INDICATION. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE MET. THIS EVENT WAS CAUSED BY DRIFT OF THE LEAD / LAG CARD GAIN. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-2-STP-201.20 (REACTOR COOLANT SYSTEM TE-432B AND TE-432D LOOP CALIBRATION AND FUNCTIONAL TEST), THE TI-432D INSTRUMENTATION CHANNEL WAS DECLARED OPERABLE AT 0540 ON 12/28/83.

FORM 98 LER SCSS DATA 07-22-86 NSIC EVENT DATE DOCKET YEAR LER NUMBER REVISION DCS NUMBER 364 1983 066 0 8401303306 188451 12/31/83 DOCKET:364 FARLEY 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: BESS FACILITY OPERATOR: ALABAMA POWER CO.

SYMBOL: APC ABSTRACT AT 0817 ON 12/31/83, THE INSTRUMENTATION CHANNEL ASSOCIATED WITH FLOW TRANSMITTER FT-476 (LOOP 1 FEED FLOW) WAS DECLARED IN0PERABLE DUE TO ERR 0NE0US INDICATION. TECH SPEC 3.3.1, IN PART, REQUIRES THIS CHANNEL TO BE OPERABLE. TECH SPEC 3.3.1 ACTION STATEMENT REQUIREMENTS WERE NET. THIS EVENT WAS CAUSED BY DRIFT OF THE SQUARE ROOT EXTRACTION CARD. THE CARD WAS REPLACED AND FOLLOWING SATISFACTORY PERFORMANCE OF FNP-2-STP-215.1 (MAIN FEEDWATER FLOW FT-476 CALIBRATION AND FUNCTIONAL TEST N2C22FT0476), THE FT-476 INSTRUMENTATION CHANNEL WAS RETURNED TO SERVICE AT 1230 ON 1/1/84.

99 LER SCSS DATA 07-22-86 FORM NSIC EVENT DATE DOCKET YEAR LER NUMBER REVISION DCS NUMBER 369 1981 051 0 8107200392 167771 04/12/81 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT t TRAIN A 0F THE SOLID STATE PROTECTION SYSTEM WAS DECLARED IN0PERABLE BECAUSE SEVERAL ERR 0NE0US PERMISSIVE BLOCK INDICATIONS APPEARED ON THE CONTROL ROOM STATUS PANEL. TWO CIRCUIT BOARDS WERE FOUND TO BE DEFECTIVE DUE TO TEMPERATURE RELATED DEGRADATION. THE DEFECTIVE CIRCUIT BOARDS WERE REPLACED AND THE SYSTEM WAS RETURNED TO SERVICE.

I

100 LER SCSS DATA 07-22-86 FORM NSIC EVENT DATE DOCKET YEAR LER NUMBER REVISION DCS NUMBER 369 1981 125 0 8109080214 168649 07/28/81 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC ABSTRACT DURING THE PERIODIC FUNCTIONAL TEST OF THE REACTOR PROTECTIVE SYSTEM, THE CHANNEL 4 OVERTEMPERATURE DELTA T INSTRUMENTATION WAS FOUND TO BE THE OUT OF CALIBRATION AND SUBSEQUENTLY DECLARED IN0PERABLE PER T.S.

CAUSE WAS DETERMINED TO BE A FAULTY CARD IN THE OVERTEMPERATURE DELTA T SOLID STATE CIRCUITRY. CHANNEL 4 WAS PLACED IN THE TRIPPED CONDITION (2 OUT OF 3 COINCIDENCE LOGIC VICE NORMAL 2 OUT OF 4 COINCIDENCE LOGIC), A REPLACEMENT CARD INSTALLED, AND CHANNEL FUNCTIONABILITY VERIFIED.

FORM 101 LER SCSS DATA 07-22-86 NSIC EVENT DATE DOCKET YEAR LER NUMBER REVISION DCS NUMBER 369 1981 172 0 8112140310 182058 10/31/81 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT WHILE IN MODE 3 IN PREPARATION FOR UNIT STARTUP, A LOOP T(AVG)

DEVIATION, A LOOP D/T DEVIATION, AND A LO-L0 T(AVG) ALERT ALARMED ON THE CONTROL ROOM ANNUNCIATOR, AND WERE ACCOMPANIED BY A "PCS POWER SUPPLY FAILURE PROTECTION CABINET 1" INDICATION. SUBSEQUENTLY, PURSUANT TO TECH SPEC 3.3.1 WHICH IS REPORTABLE PER TECH SPEC 6.9.1.13(B) AND SIMILAR TO R0 369/81-125, THE LOOP 'A' OVERTEMPERATURE DELTA T AND OVERPOWER DELTA T REACTOR PROTECTION CHANNELS WERE MANUALLY TRIPPED. TROUBLESHOOTING ISOLATED T(COLD) FAILURE TO AN NRA CARD IN THE WESTINGHOUSE PROCESS 7300 CONTROL SYSTEM. THE LOOP 'A' ASSOCIATED CHANNELS WERE TRIPPED, PUTTING THE PROTECTION COINCIDENCE LOGIC IN A MORE CONSERVATIVE 1 OUT OF 3 CONDITION. THE MALFUNCTIONING NRA CARD WAS REPLACED AND THE LOOP 'A' TEMPERATURE INSTRUMENTATION WAS CALIBRATED AND DECLARED OPERABLE.

FORM 102 LER SCSS DATA 07-22-86

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE

.369 1983 057 0 8308260092 185354 07/16/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

' SYMBOL: DPC ABSTRACT WHILE IN MODE 1, ON 7/16/83 AND AGAIN ON 7/21/83, REACTOR COOLANT LOOP B OVERTEMPERATURE DELTA T INSTRUMENT LOOP WAS DECLARED IN0FERABLE WHEN THE CONTROL ROOM INDICATOR WAS FOUND READING HIGHER THAN OTHER CHANNELS DURING ROUTINE SURVEILLANCE. THIS VIOLATES TECH SPEC 3.3.1 WHICH IS REPORTABLE PER TECH SPEC 6.9.11(B) AND SIMILAR TO R0'S 369/81-125 AND 82-18. THE FAILURE MODE OF LOOP B 0F DELTA T WAS CONSERVATIVE (HIGH) AND THIS CHANNEL WOULD HAVE TRIPPED EARLIER THAN THE OTHERS IN ANY EYENT. OTHER CHANNELS WERE OPERABLE. THE CHANNEL WAS PLACED IN THE TRIPPED CONDITION PER ACTION STATEMENT. THE CABINET CONTAINING THE CIRCUITRY WAS HOT DURING INCREASES IN CONTROL ROOM TEMPERATURE DUE TO CHILLER PROBLEMS (SEE R0-369/83-56). PROBLEM ATTRIBUTED TO HEAT SENSITIVE CIRCUIT CARDS IN THE 7300 PROCESS CONTROL SYSTEM EVEN THOUGH CONTROL ROOM TEMPERATURE DID NOT EXCEED LIMITS.

INDICATOR RETURNED TO NORMAL UPON COOLING PREVENTING IDENTIFICATION OF DEFECTIVE CARD. METHODOLOGY TO IDENTIFY AND CORRECT CARD IN EVENT OF ANOTHER MALFUNCTION ESTABLISHED.

FORM 103 LER SCSS DATA 07-22-86

- - - - +************************,****************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 369 1983 090 0 8311040019 186978 09/24/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL EN3INEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC ABSTRACT WHILE IN MODE 1, AFTER A " LOOP DELTA T DEVIATION" ALARM WAS RECEIVED IN THE CONTROL ROOM, LOOP D OVERPOWER DELTA T AND OVERTEMPERATURE DELTA T WAS DECLARED IN0PERABLE (LOOP D DELTA T IN ICATOR READ "ZER0"). THIS CONSTITUTES A DEGRADATION OF REACTOR TRIP SYSTEM INSTRUMENTATION (TECH SPEC 3.3.1; TABLE 3.3-1 FUNCTIONAL UNITS 7 AND

8) WHICH IS REPORTABLE PURSUANT TO TECH SPEC 6.9.1.11(B) AND SIMILAR TO R0'S 369/81-125, 82-18 AND 83-57. FAILURE OF THIS LOOP WAS IN THE CONSERVATIVE DIRECTION AND DID NOT AFFECT THE OTHER 3 LOOPS (A, B, AND C). THIS IS ATTRIBUTED TO COMPONENT FAILURE BECAUSE A LEAD / LAG AMPLIFIER (NLL) CARD (TY-441K) FAILED IN THE PROCESS CONTROL SYSTEM.

THE IN0PERABLE LOOP WAS PLACED IN THE " TRIP" CONDITION WITHIN 1 HOUR

PER TECH SPEC ACTION STATEMENTS, CHANGING THE TRIP LOGIC FROM 2 OUT OF 4 TO A MORE CONSERVATIVE 1 OUT OF 3. THE BAD CARD WAS REPLACED, THE LOOP CALIBRATED AND DECLARED OPERABLE.

FORM 104 LER SCSS DATA 07-22-86

- - - - +**************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NS: . EVENT DATE 369 1983 095 0 8311210022 167155 09/30/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE r ARCHITECTURAL ENGINEER: DUKE

~

rACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT WHILE IN MODE 1, STEAM GENERATOR 'A' WATER LEVEL HIGH-HIGH CHANNEL III WAS DECLARED IN0PERABLE AFTER "S/G HI-HI LEVEL ALERT", "PCS PWR SUPPLY FAILURE PROT. CAB. 3", AND "S/G A LB518A HI-HI LEVEL" ALARMS WERE RECEIVED IN THE CONTROL ROOM. THIS CONSTITUTES A DEGRADATION OF ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (TECH SPEC 3.3.2; TABLE 3.3-3 FUNCTIONAL UNIT 5.B) WHICH IS REPORTABLE PURSUANT TO TECH SPEC 6.9.1.11(B). CHANNEL III WAS NOT USED FOR CONTROL PURPOSES AND ITS FAILURE WAS IN THE CONSERVATIVE DIRECTION. THE OTHER 2 CHANNELS (II AND IV) WERE NOT AFFECTED. This IS ATTRIBUTED TO COMPONENT FAILURE BECAUSE THE SIGNAL COMPARATOR (NAL) CARD IN THE S/G A WATER LEVEL HIGH-HIGH CHANNEL III SECTION OF THE PROCESS CONTROL CABINET FAILED UNDER NORMAL OPERATING CONDITIONS. THE CARDS FAILURE APPEARS TO BE RAND 0M. THE CHANNEL WAS PLACED IN THE " TRIP" CONDITION.

PER TECH SPEC ACTION STATEMENT, CHANGING THE TRIP LOGIC FROM 2 OUT OF 3 TO 1 OUT OF 2. THE FAILED CARD WAS REPLACED AND THE CHANNEL RETURNED TO SERVICE.

FORM 105 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUM3ER NSIC EVENT DATE 369 1983 103 0 8311210290 186936 10/12/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:kE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC ABSTRACT ON 10/12/83, WHILE IN MODE 1, REACTOR C0OLANT SYSTEM LOOP B OVERTEMPERATURE DELTA T WAS DECLARED IN0PERABLE WHEN THE LOOP INDICATOR EXCEEDED THE PLUS OR MINUS 2% TOLERANCE ALLOWED DURING SEMI-DAILY SURVEILLANCE. THIS CONSTITUTES A DEGRADATION OF REACTOR

TRIP SYSTEM INSTRUMENTATION (TECH SPEC 3.3.1, TABLE 3.3-1) WHICH IS l REPORTABLE PER TECH SPEC 6.9.1.11(B) AND IS SIMILAR TO R0'S 369/81-125, 82-18, 83-57, 83-90 AND 370/83-60. LOOPS A, C AND D REMAINED OPERABLE. THIS FAILURE IS ATTRIBUTED TO THE MALFUNCTION OF HEAT SENSITIVE COMP 0NENTS (M0TOROLA ZENER DIODES AND WESTINGHOUSE OP-AMPS) ON A PRINTED CIRCUIT CARD IN THE PROCESS CONTROL SYSTEM. THE IN0PERABLE LOOP WAS PLACED IN THE " TRIP" CONDITION WITHIN 1 HOUR AS REQUIRED BY TECH SPEC ACTION STATEMENTS. HEAT SENSITIVIE COMP 0NENTS WERE REPLACED ON THE CIRCUIT CARD AND THE SYSTEM WAS DECLARED OPERABLE AFTER APPROPRIATE SURVEILLANCE.

FORM 106 LER SCSS DATA 07-22-86 NSIC EVENT DATE DOCKET YEAR LER NUMBER REVISION DCS NUMBER 369 1983 104 0 8311210294 186979 10/20/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC ABSTRACT WHILE IN MODE 1, REACTOR COOLANT SYSTEM LOOP B OVERTEMPERATURE DELTA T WAS DECLARED IN0PERABLE WHEN THE LOOP INDICATOR EXCEEDED THE PLUS OR MINUS 2% TOLERANCE ALLOWED DURING SEMI-DAILY SURVEILLANCE. THIS CONSTITUTES A CEGRADATION OF REACTOR TRIP SYSTEM INSTRUMENTATION (TECH SPEC 3.3.I, TABLE 3.3-1) WHICH IS REPORTABLE PER TECH SPEC 6.9.1.11(B) AND IS SIMILAR TO R0'S 369/81-125, 82-18, 83-57, 83-90,83-103, AND 370/83-60. LOOPS A, C, AND D REMAINED OPERABLE. THIS FAILURE IS ATTRIBUTED TO THE MALFUNCTION OF HEAT SENSITIVE COMPONENTS (NATIONAL SEMICONDUCTOR OP-AMP) ON A PRINTED CIRCUIT CARD IN THE PROCESS CONTROL SYSTEM. THE IN0PERABLE LOOP WAS PLACED IN THE " TRIP" CONDITION WITHIN 1 HOUR AS REQUIRED BY TECH SPEC ACTION STATEMENTS.

HEAT SENSITIVE COMP 0NENTS WERE REPLACED ON THE CIRCUIT CARD AND THE SYSTEM WAS DECLARED OPERABLE AFTER APPROPRIATE SURVEILLANCE.

FORM 107 LER SCSS DATA 07-22-86 D0CKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 369 1983 108 0 8312160124 187817 11/03/83 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT

1

-WHILE IN MODE 1, DURING PERIODIC TESTING OF THE SOLID STATE PROTECTION SYSTEM TRAIN "A", A BAD UNIVERSAL LOGIC CARD WAS FOUND WHICH WOULD NOT HAVE PROVIDED FEEDWATER ISOLATION OF A LOW T(AVE) SIGNAL AFTER A REACTOR TRIP (FROM TRAIN A). THIS CONSTITUTES A DEGRADATION OF THE REACTOR TRIP SYSTEM INSTRUMENTATION (TECH SPEC 3.3.1) AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (TECH SPEC 3.3.2)

WHICH IS REPORTABLE PURSUANT TO TECH SPEC 6.9.1.11(B). THE TRAIN B LOGIC CIRCUITRY WAS SATISFACTORILY TESTED AND WOULD HAVE BEEN CAPABLE OF PROVIDING FEEDWATER ISOLATION IF REQUIRED. THIS IS ATTRIBUTED TO COMPONENT FAILURE DUE TO A FAILED TWO-0UT-0F-FOUR LOGIC CIRCUIT FOUND ON A UNIVERSAL LOGIC CARD. THE UNIVERSAL LOGIC CARD WAS REPLACED, SSPS 1 TRAIN A SATISFACTORILY TESTED AND DECLARED OPERABLE.

TROUBLESHOOTING, REPAIR, AND TESTING EXCEEDED THE TWO HOUR LIMIT THAT TECH SPEC ACTION STATEMENTS PERMIT FOR 1 TRAIN TO BE BYPASSED, THEREFORE LOAD REDUCTION WAS COMMENCED UNTIL TRAIN A WAS DECLARED OPERABLE (I HOUR LATER AT 93% POWER).

FORM 108 LER.SCSS DATA 07-22-86 +

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 369 1984 002 0 8403130332 189405 01/30/84 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC ABSTRACT POWER LEVEL - 094%. ON JANUARY 30, 1984 A UNIT 1 REACTOR TRIP WAS I

INITIATED BY THE REACTOR PROTECTION SYSTEM ON TWO-0UT-0F-FOUR OVERTEMPERATURE DELTA T (0 T DELTA T) SIGNAL. THE TRIP OCCURRED FROM A VOLTAGE SPIKE IN LOOP C WHILE LOOP A WAS IN THE TEST (TRIP)

CONDITION FOR MAINTENANCE. UNIT 1 WAS IN MODE 1 AT 94% POWER AT THE TIME OF THE TRIP. THIS EVENT IS ATTRIBUTED TO THE FAILURE OF A LEAD / LAG CARD IN LOOP C DELTA T/TAVG OF THE FR0 CESS CONTROL SYSTEM.

THE REACTOR TRIPPED AS DESIGNED AND N0 AN0MALIES ARISING FROM THE TRANSIENT OCCURRED. THE FAULTY CARD WAS REPLACED AND THE LOOP RETURNED TO SERVICE.

' FORM 109 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 369 1984 015 0 8406130072 190376 04/26/84 i ********************************************************************

\

DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

._ - - .~. - . ... . . .

, - e. <-

l l

SYMBOL: DPC ABSTRACT

' POWER LEVEL - 000%. ON APR 26, 1984 THE " PROCEDURE FOR FULL LENGTH ROD CONTROL CLUSTER ASSEMBLY DROP TIMING" WAS PERFORMED TO VERIFY DIGITAL R0D POSITION INDICATION (DRPI) ACCURACY AND PROPER OPERATION OF THE CONTROL ROD DRIVE SYSTEM. AT 0830 THE CONTROL OPERATOR OPENED THS

, REACTOR TRIP BREAKERS IN ACCORDANCE WITH THE ACTION STATEMENT OF TECH SPEC 3.1.3.3 WHEN ROD B-12, IN SHUTDOWN BANK A, FAILED TO INDICATE ITS CORRECT POSITION. UNIT 1 WAS IN MODE 3 WITH THE CONTROL AND SHUTDOWN BANKS INSERTED EXCEPT SHUTDOWN BANK A, WHICH WAS 18 STEPS WITHDRAWN.

4 THIS EVENT IS ATTRIBUTED TO COMPONENT FAILURE BECAUSE A CIRCUIT ON THE DETECTOR / ENCODER CARD FAILED. THIS CARD FAILURE GAVE AN INCORRECT INDICATION FOR THE LOCATION OF ROD B-12. THE FAULTY CARD WAS REPLACED WITH A SPARE AND THE DROP TIMING PROCEDURE WAS SUCCESSFULLY COMPLETED. THE REACTOR REMAINED SUBCRITICAL THROUGHOUT THE EVENT.

FORM 110 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 369 1984 018 1 8504040462 196608 06/04/84 j DOCKET:369 MCGUIRE I TYPE:PWR

REGION: 2 NSSS:WE

! ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC i- ABSTRACT POWER LEVEL - 100%. ON 6-4-84 AT APPR0X 2000, THE TRAIN B CHILLER OF THE CONTROL AREA VENTILATION SYSTEM TRIPPED DUE TO LOW OIL LEVEL AND WAS DECLARED INOPERABLE. TRAIN A 0F VC HAD BEEN PREVIOUSLY DECLARED INOPERABLE BECAUSE OF MAINTENANCE WORK. THE IN0PERABILITY OF BOTH TRAINS OF VC, WHILE A UNIT IS ON-LINE, IS PROHIBITED BY TECH SPEC

3.7.6. ACCORDINGLY, AT 2205 THE CONTROL OPERATORS STARTED TO REDUCE POWER ON UNITS 1 AND 2 AS REQUIRED BY TECH SPEC 3.0.3. UNITS 1 AND 2 WERE IN MODE 1 AT 100% POWER AT THE TIME OF THIS EVENT. AT APPR0X

, 2230, 5 GALS OF OIL WERE-ADDED TO THE CHILLER AND THE CHILLER >

RESTARTED. WITH VC TRAIN B THEN OPERABLE, THE CONTROL OPERATORS

STOPPED REDUCING POWER WITH EACH UNIT HAVING REACHED 97% POWER. TRAIN B 0F VC WAS DECLARED OPERABLE AT 2255. THE UNITS WERE RETURNED TO l 100% POWER AT 2312. THIS EVENT IS ATTRIBUTED TO UNUSUAL SERVICE

CONDITIONS, DUE TO THE COOLING LOAD OF THE CONTROL ROOM AREA BEING INSUFFICIENT TO FULLY LOAD THE TRAIN B CHILLER. DUKE POWER IS CONTINUING ITS REVIEW 0F THIS PROBLEM TO DETERMINE ANY FURTHER CORRECTIVE ACTIONS. IN ADDITION, THE FAILURES OF PRINTED CIRCUIT CARDS IN THE PROCESS CONTROL SYSTEM (PCS) CABINETS, WHICH HAVE l

~

OCCURRED IN THIS AND OTHER EVENTS INVOLVING OVERHEATING IN THE PCS CABINETS, HAVE BEEN EXAMINED. USE OF HEAT SINKS AND IMPROVED COOLING IN THE PCS CABINETS IS EXPECTED TO ALLEVIATE THE PROBLEMS.

I i

FORM 111 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 369 1984 020 0 8407190365 190684 06/06/84 DOCKET:369 MCGUIRE 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT POWER LEVEL - 090%. ON JUN 6, 1984, UNIT 1 REACTOR TRIPPED AT 1756 WHEN A SPURIOUS UNDERV0LTAGE RELAY ACTUATION DEENERGIZED ONE OF THE REACTOR COOLANT (NC) PUMPS. THF RELAY WHICH PROTECTS 7KV BUS ITC FROM UNDERVOLTAGE OPERATION BY MONITORING THE NORMAL INCOMING POWER, MALFUNCTIONED AND TRIPPED THE NORMAL INCOMING POWER CIRCUIT BREAKER ITC-6. LOSS OF THIS POWER SOURCE DEENERGIZED 7KV BUS 1TC, WHICH INCLUDED NC PUMP IC. SINCE THE UNIT WAS OPERATING AT A POWER LEVEL AB0VE THE AUTOMATIC TRIP SETPOINT OF 48% (90% ACTUAL THERMAL POWER),

LOSS OF THE NC PUMP CAUSED AN IMMEDIATE REACTOR AND TURBINE TRIP. THE RELAY WAS REPLACED AND THE REACTOR RESTARTED. THE EVENT IS ATTRIBUTED TO COMPONENT MALFUNCTION SINCE THE RELAY SPURIOUSLY ACTUATED WITH NORMAL VOLTAGE ON THE NORMAL INCOMING POWER SOURCE. INITIAL TESTING INDICATED THAT THE RELAY APPEARS TO BE WORKING CORRECTLY. FURTHER TESTING OF THE RELAY WILL BE CONDUCTED TO DETERMINE THE FAILURE MECHANISM 0F THE DEVICE. THE UNDERVOLTAGE RELAY FAILED IN THE CONSERVATIVE DIRECTION, AND PLANT RESPONSE TO THE TRIP WAS WELL CONTROLLED.

FORM 112 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 370 1983 020 0 8306300114 184239 05/19/83 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT WHILE IN MODE 2, SHUTDOWN R0D BANK "A" GROUP 1 AND CONTROL R0D BANKS "A" GROUP 1 AND "C" GROUP 1 FAILED TO MOVE ON COMMAND Arl0 WERE DECLARED INOPERABLE. TPIS VIOLATES TECH SPEC 3.1.3.1 QHICH IS '

REPORTABLE PER TECH SPEC 6.9.1.11(B). THE TEMPORARY FAILURE OF THE R0D GROUPS Vf.S LIMITED TO LIFT COIL OPERATION ONLY (N0 BINDING OR ,

STICKING OF THE R0D ASSEMBLIES OCCURRED). THE R0DS WERE THEREFORE FULLY CAPABLE OF DROPPING UPON RECEIVING A MANUAL OR AUTOMATIC REACTOR TRIP SIGNAL. THE FAILED R0D GROUPS SHARE A COMMON POWER SUPPLY. THE l

l l

l 1

4 .

TROUBLE WAS TRACED TO INPUT /0UTPUT AC AMPLIFIER CARD A805 (WESTINGHOUSE 3359C65G01) LOCATED IN THE LOGIC CABINET, WHICH SENDS CURRENT COMMANDS FOR THE CONTROL ROD DRIVE MECHANISM LIFT COILS TO THE POWER CABINET. THE CARD MALFUNCTION WAS DUE TO FAILURE OF TRANSISTORS 07 AND 08 (TYPE 2N2219A). THE CARD WAS REPLACED AND ROD GROUPS DECLARED OPERABLE.

FORM 113 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 370 1983 052 0 8311010042 186469 09/19/83 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT WHILE IN MODE 1, B STEAM GENERATOR LOW STEAMLINE PRESSURE CHANNEL II WAS DECLARED INOPERABLE DUE TO RECURRING ANNUNCIATOR ALARM AND STATUS INDICATOR ALARM IN THE CONTROL ROOM. THIS CONSTITUTES A DEGRADATION OF ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (TECH SPEC 3.3.2, TABLE 3.3-3 FUNCTIONAL UNIT 1.E) WHICH IS. REPORTABLE PURSUANT TO TECH SPEC 6.9.1.11(B). FAILURE OF CHANNEL II WAS IN THE C)NSERVATIVE DIRECTION AND THE OTHER TWO CHANNELS (I AND III) WERE NOT AFFECTED. THIS INCIDENT IS ATTRIBUTED TO COMPONENT FAILURE BECAUSE THE LEAD / LAG AMPLIFIER CARD IN THE PROCESS CONTROL SYSTEM FAILED (0UTPUT " SWINGING" EXCESSIVELY) UNDER NORMAL OPERATING CONDITIONS.

THE CHANNEL WAS PLACED IN THE " TRIP" CONDITION PER TECH SPEC ACTION STATEMENT, CHANGING THE TRIP LOGIC FROM 2 OUT OF 3 TO 1 OUT OF 2. A NEW LEAD / LAG AMPLIFIER CARD WAS INSTALLED AND THE CHANNEL RETURNED TO SERVICE.

FORM 114 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 370 1983 060 0 8311170349 187017 10/10/83 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT WHILE IN MODE 1, REACTOR COOLANT SYSTEM LOOP D OVERPOWER DELTA T AND OVERTEMPERATURE DELTA T WAS DECLARED INOPERABLE WHEN A LOOP DELTA /T DEVIATION ALARM WAS RECEIVED IN THE CONTROL ROOM. (LOOP D DELTA /T

- ~ . . - , - -

1 INDICATOR READ "ZER0".) THIS CONSTITUTES A DEGRADATION OF REACTOR TRIP SYSTEM INSTRUMENTATION (TECH SPEC 3.3.1, TABLE 3.3-1) WHICH IS REPORTABLE PER TECH SPEC 6.9.1.11(B) AND IS SIMILAR TO R0'S 369/81-125, 82-18, 83-57 AND 83-90. FAILURE OF THIS LOOP WAS IN THE CONSERVATIVE DIRECTION AND DID NOT AFFECT LOOPS A, B, OR C. THIS FAILURE IS ATTRIBUTED TO FAILURE OF A LEAD / LAG AMPLIFIER (NLL) CARD IN THE PROCESS CONTROL SYSTEM. THE IN0PERABLE LOOP WAS PLACED IN THE

" TRIP" CONDITION WITHIN 1 HOUR AS REQUIRED BY TECH SPEC ACTION STATEMENTS. THE BAD NLL CARD WAS REPLACED WHICH CLEARED THE ALARM AND CORRECTED THE INDICATOR READING. REPAIRS WERE INITIATED ON THE BAD NLL CARD.

FORM 115 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 370 1984 011 0 8406070281 189678 04/23/84 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER CO.

SYMBOL: DPC ABSTRACT POWER LEVEL - 100%. ON APR 23, 1984 AT 0057, CONTROL ROOM ANNUNCIATOP

" PROCESS CONTROL SYSTEM POWER SUPPLY FAILURE PROTECTION CABINET I" ALARMED AND STATUS ALARMS INDICATED A LOSS OF CHANNEL 1, WHICH WAS BEING USED FOR STEAM GENERATOR AND PRESSURIZER CONTROL. ALL 4 STEAM GENERATOR LEVELS BEGAN INCREASING BECAUSE THE 4 FEEDWATER REGULATOR VALVES OPENED. THE CONTROL OPERATORS PLACED THE 4 SG LEVEL CONTROLS INTO " MANUAL". ONE CONTROL OPERATOR MADE ADJUSTMENTS WITH THE LEVEL CONTROLLERS TO DECREASE LEVELS IN SG'S A AND B. ANOTHER CONTROL OPEPATOR WAS ENSURING THAT THE CONTROLS WERE CHANGED FROM CHANNEL 1 TO CHANNEL 2. THEN HE BEGAN TC MAKE LEVEL ADJUSTMENTS ON SG'S C AND D.

THE ADJUSTMENTS ON SG'S C AND D WERE NOT MADE IN TIME, WHICH ALLOWED SG D TO REACH THE HI-HI LEVEL TRIP SETPOINT AT 82% AT 0100. UNIT 2 WAS IN MODE 1, AT 100% POWER, AT THE TIME OF THE TURBINE / REACTOR TRIP.

THIS EVENT IS ATTRIBUTED TO COMPONENT FAILURE DUE TO THE FAILURE OF THE MAIN POWER SUPPLY IN THE PROCESS CONTROL SYSTEM PROTECTION CABINET I. DESIGN DEFICIENCY ALSO CONTRIBUTED TO THE EVENT BECAUSE POWER TO BOTH THE MAIN AND BACK-UP POWER SUPPLIES IN EACH PROTECTION CABINET IS SUPPLIED BY 1 SUPPLY BREAKER FOR EACH CABINET. THE DEFECTIVE POWER SUPPLY WAS REPLACED WITH A SPARE, AND OPERATING PROCEDURES WERE REVISED. MODIFICATIONS WILL BE MADE TO PLACE THE BACK UP POWER SUPPLIES ON SEPARATE SUPPLY BREAKERS. THE REACTOR TRIPPED DUE TO A TURBINE TRIP AB0VE 49% POWER.

FORM 116 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE

370 1984 031 0 8501080577 192659 11/24/84 DOCKET:370 MCGUIRE 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT POWER LEVEL - 100%. ON 11-24-84, UNIT 2 TRIPPED FROM 100% POWER WHEN THE FAILURE OF A LOOP POWER SUPPLY (NLP) CARD IN THE PROCESS CONTROL SYSTEM (PCS) CAUSED A DOWNWARD SPIKE IN ONE CHANNEL OF THE OVERTEMPERATURE DIFFERENTIAL TEMPERATURE (0 TDT) SETPOINT CIRCUIT. THE TRIP OCCURRED BECAUSE A SECOND CHANNEL OF THE OTDT CIRCUIT WAS ALREADY IN THE TEST (TRIPPED) MODE, DUE TO AN INOPERABLE INSTRUMENT, WITH ONE CHANNEL IN THE TEST MODE, COMBINED WITH THE SPIKE IN ANOTHER CHANNEL, THE 2 OUT OF 4 LOGIC REQUIRED FOR A REACTOR TRIP SIGNAL WAS SATISFIED. THE CAUSE OF THE FAILURE OF THE NLP CARD IS SUSPECTED TO BE OVERHEATING IN THE PCS CABINEi. FAILURES OF PCS CARDS IN THE PAST HAVE PROMPTED SEVERAL CORRECTIVE ACTIONS TO BE COMPLETED AND/0R INITIATED. THESE ACTIONS ADDRESS IMPROVED PCS CABINET VENTILATION AND USE OF HEAT SINKS TO AVOID OVERHEATING, AND IMPLEMENTATION OF PROGRAMS TO IMPROVE CARD RELIABILITY. THE INTERMITTENT FAILURE OF POWER RANGE NUCLEAR INSTRUMENTATION 41, WHICH CAUSED THE SECOND CHANNEL OF THE OTDT CIRCUIT TO BE IN THE TEST MODE, IS SUSPECTED TO BE CAUSED BY A P00R CONNECTION IN A CABLE, CONNECTOR, OR DETECTOR. THE CABLES, CONNECTORS, OR DETECTOR WILL BE REPAIRED OR REPLACED.

FORM 117 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1982 018 0 8211100299 178958 10/08/82 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE APCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS CO.

SYMBOL: SCC ABSTRACT WHILE PERFORMING SURVEILLANCE TEST PROCEDURE (STP) 303.006 ON THE B STEAM GENERATOR MAIN STEAM LINE PRESSHRE INSTRUMENTATION, A PRESSURE BISTABLE FAILED TO CHANGE OUTPUT STATE WHEN VARYING THE INPUT ABOUT THE TRIP AND RESET POINTS. INVESTIGATION REVEALED THAT A PRINTED CIRCUIT LOGIC CARD HAD FAILED. THIS CARD WAS REPLACED, AND THE SURVEILLANCE TEST PROCEDURE WAS PERFORMED SATISFACTORY. THE SYSTEM WAS RETURNED TO OPERABLE STATUS ON OCTOBER 8, 1982. THE LICENSEE PLANS NO FURTHER ACTIONS FOR THIS EVENT.

FORM 118 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1982 053 0 8301170104 181595 12/06/82 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS C0.

SYMBOL: SCC ABSTRACT ON DECEMBER 6, 1982 WITH THE PLANT IN MODE 1, SPURIOUS HIGH AND LOW METER INDICATIONS WERE OBSERVED ON THE MAIN CONTROL BOARD INDICATOR (FI-494) FOR STEAM GENERATOR 'C' STEAM FLOW. THERE WERE NO ADVERSE CONSEQUENCES AS THE MINIMUM NUMBER OF CHANNELS PER LOOP IDENTIFIED IN TECH SPEC TABLES 3.3-1 AND 3.3-3 REMAINED OPERABLE DURING REPAIR OF THE TRIFPED CHANNEL. THE EVENT IS ATTRIBUTED TO THE FAILURE OF CIRCUIT BOARD FY-494. THE CHANNEL WAS PLACED IN THE TRIPPED CONDITION WITHIN ONE HOUR. THE DEFECTIVE CIRCUIT BOARD WAS REPLACED UPON IDENTIFICATION OF FAILURE, AND THE CHANNEL WAS RETURNED TO CPERABLE STATUS ON DECEMBER 7, 1982, UPON COMPLETION OF A SATISFACTORY SURVEILLANCE TEST. N0 FURTHER ACTION IS PLANNED BY THE LICENSEE.

FORM 119 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 003 0 8400000000 188715 01/08/83 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE i ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS CO.

SYMBOL: SCC ABSTRACT ON JAN. 8, 1983, MAIN CONTROL BOARD INDICATION FOR STEAM GENERATOR "A" FEEDWATER FLOW (IFI-477) FAILED LOW. ON JAN. 20, 1983, IFI-477 FAILED 20% BELOW REDUNDANT CHANNELS. IFI-477 FEEDS REACTOR PROTECTION CIRCUITRY FOR A TRIP ON STEAM FLOW - FEEDWATER FLOW MISMATCH CONCIDENT WITH A LOW STEAM GENERATOR WATER LEVEL. THERE WERE N0 ADVERSE CONSEQUENCES AS THE INSTRUMENT FAILED IN THE SAFE DIRECTION, AND A REACTOR TRIP WOULD STILL HAVE OCCURRED UPON SETPOINT ACTUATION.

THE JAN. 8, 1983 EVENT IS ATTRIBUTED TO HIGH INSTRUMENT CABINET TEMPERATURE. THE POWER SUPPLY CARD WAS REPLACED. THE JAN. 20, 1983 EVENT IS ATTRIBUTED TO INSTRUMENT URIFT OF THE CARD REPLACED FOR THE JAN. 8, 1983 EVENT. THE CARD WAS RECALIBRATED. RELAY ROOM VENTILATION IS TO BE MODIFIED FOR MORE EFFECTIVE COOLING OF THE INSTRUMENT CABINETS AND COMPONENT 3.

FORM 120 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 012 0 8303180241 182009 02/10/83 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS CO.

SYMBOL: SCC ABSTRACT ON FEBRUARY 10, 1983, WITH THE PLANT IN MODE 1. THE FLOW INDICATOR FOR STEAM GENERATOR "A" FEEDWATER FLOW (FI-477) FAILED LOW AS COMPARED TO REDUNDANT CHANNELS. THE CHANNEL WAS DECLARED INOPERABLE, AND THE BISTABLES WERE PLACED IN THE TRIPPED CONDITION. THERE WERE NO ADVERSE CONSEQUENCES AS THE INSTRUMENT FAILED IN THE SAFE DIRECTION, AND A REACTOR TRIP WOULD STILL HAVE OCCURRED UPON SETP0 INT ACTUATION. THE CAUSE OF THIS OCCURRENCE IS ATTRIBUTED TO THE FAILURE OF POWER SUPPLY FQY-477. THE CIRCUIT BOARD WAS REPLACED AND THE CHANNEL RETURNED TO OPERABLE STATUS ON FEBRUARY 11, 1983. THE LICENSEE PLANS N0 ADDITIONAL ACTION IN REGARDS TO THIS EVENT OTHER THAN THE NORMAL SURVEILLANCE TESTING.

FORM 121 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 056 0 8307120081 184286 06/06/83 DOCKET:395 SUKMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS C0.

SYMB0'.: SCC ABSTRACT ON JUNE 6, 1983, WITH THE PLANT IN MODE 1 FEEDWATER FLOW TRANSMITTER (FT-486) B STEAM GENERATOR FAILED LOW. THE INOPERABLE CHANNEL WAS PLACED IN THE TRIPPED POSITION WITHIN ONE (1) HOUR AS REQUIRED BY ACTION STATEMENT 6 0F TABLE 3.3-1 TECH SPEC 3.3.1, " REACTOR TRIP SYSTEM INSTRUMENTATION". THE MINIMUM NUMBER OF CHANNELS AS REQUIRED BY THE AFOREMENTIONED TECH SPEC, ITEM 14, TABLE 3.3-1 WAS SATISFIED.

THE FAILURE OF FT-486 WAS DUE TO THE FAILURE OF THE POWER SUPPLY CARD.

THE CARD WAS REPLACED, THE APPLICABLE SURVEILLANCE TEST PROCEDURE SATISFACTORILY PERFORMED, AND THE UNIT DECLARED OPERABLE ON JUNE 6, 1983. NO ADDITIONAL CORRECTIVE ACTION IS PLANNED OTHER THAN REQUIRED SURVEILLANCE TESTING.

FORM 122 LER SCSS DATA 07-22-86

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 077 0 8308110112 185319 07/05/83 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS CO.

SYMBOL: SCC ABSTRACT ON JULY 5, 1983, AT 1745 HOURS, WITH THE PLANT IN MODE 1, TRAIN 'B' SOLID STATE PROTECTION SYSTEM (SSPS) WAS DECLARED INOPERABLE WHEN A MALFUNCTION ON STEAM GENERATOR 'A' L0 LO WATER LEVEL LOGIC WAS DETECTED DURING THE PERFORMANCE OF A SURVEILLANCE TEST. ALL OTHER FUNCTIONS OF TRAIN 'B' SSPS WERE FUNCTIONAL. THE PLANT WAS IN COMPLIANCE WITH ACTION STATEMENT 8 0F TECH SPEC 3.3.1 (TABLE 3.3-1, ITEM 21), AND THE REDUNDANT TRAIN WAS OPERABLE. THE CAUSE OF THE FAILURE OF TRAIN 'B' SSPS IS ATTRIBUTED TO A DEFECTIVE CIRCUIT BOARD IN THE LOGIC CIRCUITRY. THE BOARD WAS REPLACED, AND THE APPROPRIATE SURVEILLANCE TEST SATISFACTORILY PERFORMED. THE TRAIN 'B' SSPS WAS RETURNED TO OPERABLE STATUS ON JULY 5, 1983, AT 2110 HOURS. NO FURTHER ACTION IS PLANNED OTHER THAN NORMAL SURVEILLANCE TESTING.

FORM 123 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 112 0 8310310234 186725 09/23/83

- - - - u DOCKET:395 SUMMER I TYPE:PWR REGION: 2 NSSS:WE APCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS C0.

SYMBOL: SCC ABSTRACT ON SEPTEMBER 23, 1983, AT 1515 HOURS, WITH THE PLANT IN MODE 1, TFE SIGNAL COMPARATOR (IPB-95381) ASSOCIATED WITH CONTAINMENT PRESSURE TRANSMITTER IPT-953 WAS DECLARED INOPERABLE WHEN THE BISTABLE TRIPPED AND WOULD NOT RESET. THE BISTABLES ASSOCIATED WITH THE CHANNEL WERE TRIPPED AND VERIFIED IN ACCORDANCE WITH ACTION STATEMENT B 0F TECH SPEC 3.3.2. THERE WERE N0 ADVERSE CONSEQUENCES BECAUSE THE REDUNDANT PRESSURE INSTRUMENTS WERE OPERABLE. THE CAUSE OF THE CONTAINMENT PRESSURE INSTRUMENT FAILURE IS ATTRIBUTED TO A DEFECTIVE BISTABLE ELECTRONIC CARD. THE CARD WAS REPLACED, THE APPROPRIATE OPERATIOFAL TEST PERFORMED, AND THE CHANNEL WAS RETURNED TO OPERABLE STATUS AT 1850 HOURS ON SEPTEMBER 23, 1983. NO FURTHER ACTION IS PLANNED IN REGARD TO THIS EVENT.

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FORM 124 LER SCSS DATA 07-22-86 1 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 114 0 8310310392 187028 10/04/83 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS C0.

SYMBOL: SCC ABSTRACT ON OCTOBER 4, 1983, WITH THE PLANT IN MODE 1, STEAM GENERATOR (SG) "C" PRESSURE INDICATOR (PI) 494 WAS DECLARED IN0PERABLE DUE TO SPIKING CAUSING VARIOUS ALARMS AND THE BISTABLE TO ACTUATE. THERE WERE NO ADVERSE CONSEQUENCES DUE TO THIS EVENT. THE LICENSEE COMPLIED WITH ACTION STATEMENT 15, TABLE 3.3-3, OF TECH SPEC 3.3.2, " ENGINEERING SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION," AND ACTION STATEMENT (A) 0F TECH SPEC 3.3.3.6, " ACCIDENT MONITORING INSTRUMENTATION." THE CAUSE OF THE SPIKING WAS TRACED TO THE ELECTRONIC LOOP POWER SUPPLY CARD. THE BAD CARD WAS REPLACED AND THE NEW CARD WAS CALIBRATED. THE APPLICABLE SURVEILLANCE TEST WAS PERFORMED AND THE SYSTEM WAS DECLARED OPERABLE. TOTAL INOPERABLE TIME WAS FIVE HOURS AND FIFTY MINUTES. THE LICENSEE PLANS NO ADDITIONAL CORRECTIVE ACTION AS A RESULT OF THIS EVENT.

FORM 125 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1983 131 0 8312280187 187972 11/24/83 00CKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS C0.

SYMBOL: SCC ABSTRACT ON NOV. 24, 1983, WITH THE PLANT IN MODE 3. THE PRESSURIZER PRESSURE (LOW) TRIP BISTABLE PB-455D BEGAN TO ALTERNATELY TRIP AND RESET. THE EVENT OCCURRED DURING C00LDOWN FOR AN EDDY CURRENT OUTAGE. THE BISTABLE MALFUNCTION BEGAN WITH THE REACTOR COOLANT SYSTEM PRESSURE BETWEEN 1850 PSIG AND 1700 PSIG. THE CHANNEL WAS DECLARED INOPERABLE AT 0915 HOURS AND TRIPPED AT 0930 HOURS IN ACCORDANCE WITH ACTION STATEMENT 15 0F TECH SPEC 3.3.2. THERE WERE NO ADVERSE CONSEQUENCES FROM THIS EVENT. THE CAUSE OF THIS EVENT IS ATTRIBUTED TO AN ELECTRONIC COMPONENT FAILURE. THE DEFECTIVE CIRCUIT BOARD WAS REPLACED AND THE INSTRUMENTATION CHANNEL RETURNED TO OPERABLE STATUS ON DECEMBER 3,1983, UPON THE SATISFACTORY COMPLETION OF A CALIBRATION AND OPERATIONAL SURVEILLANCE TEST. NO FURTHER ACTION IS PLANNED

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l REGARDING THIS ISOLATED OCCURRENCE.

FORM 126 LER SCSS DATA 07-22-86

- - - - w********************************************

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1984 008 0 8403090052 189076 02/07/84 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS CO.

SYMBOL: SCC ABSTRACT POWER LEVEL - 100%. ON FEBRUARY 7, 1984, THE REACTOR TRIPPED FROM NORMAL OPERATIONS AT 100% BECUASE OF A LOW-LOW LEVEL IN STEAM GENERATOR "B". A MAIN FEEDWATER REGULATING VALVE PARTIALLY CLOSED WHEN AN ELECTRONIC CARD (FCY-488) IN ITS CONTROL CIRCUITRY FAILED.

THE PARTIALLY CLOSED FEEDWATER REGULATING VALVE SUBSEQUENTLY CAUSED THE STEAM GENERATOR "B" LOW-LOW LEVEL CONDITION. THE PLANT RESPONDED AS EXPECTED DURING THE TRANSIENT. A POST TRIP REVIEW WAS SATISFACTORILY CONDUCTED AND PLANS WERE MADE TO RESTART THE REACTOR.

FORM 127 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 395 1985 011 0 8505290446 194849 04/18/85 DOCKET:395 SUMMER 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: GLBT FACILITY OPERATOR: SOUTH CAROLINA ELECTRIC & GAS CO.

SYMBOL: SCC ABSTRACT POWER LEVEL - 100%. AT APPROXIMATELY 0147 HOURS ON APRIL 18, 1985, CONTROL RODS POWERED BY R00 CONTROL SYSTEM CABINET 2AC FAILED TO MOVE.

THE AFFECTED RODS WERE DECLARED IN0PERABLE, AND THE PLANT ENTERED ACTION STATEMENT (B) 0F TECH SPEC 3.1.3.1, " MOVABLE CONTROL ASSEMBLIES." DURING SYSTEM TROUBLESHOOTING, AN INADVERTENT R0D DROP OCCURRED JUST PRIOR TO STARTING A CONTROLLED SHUTDOWN OF THE PLANT.

THE R0D DROP CAUSED A REACTOR TRIP FROM 100% POWER AT 0629 HOURS ON POWER RANGE NEGATIVE RATE. THERE WERE N0 ADVERSE CONSEQUENCES FROM THIS EVENT. THE REACTOR PROTECTION SYSTEM FUNCTIONED PER DESIGN. DUE TO A PREVIOUS STEAM GENERATOR TUBE LEAK, THE STEAM DUMPED FROM THE AFFECTED STEAM LINE TO ATMOSPHERE AND FROM THE TURBINE-DRIVEN EMERGENCY FEEDEWATER PUMP EXHAUST RESULTED IN AN UNM0NITORED RELEASE.

THE RELEASE WAS CONSERVATIVELY CALCULATED TO BE A SMALL FRACTION OF THE ALLOWABLE RELEASE LIMITS. THE R0D CONTROL SYSTEM FAILURE WAS

DETERMINED TO BE A DEFECTIVE SLAVE CYCLER COUNTER CARD. THE CARD WAS REPLACED AND A PLANT RESTART MADE AFTER A FORCED OUTAGE OF 19.3 HOURS.

TO PREVENT A P0TENTIAL RECURRENCE, THE LICENSEE HAS ESTABLISHED A PREVENTIVE MAINTENANCE PROGRAM FOR THE R0D CONTROL SYSTEM CABINETS.

FORM 128 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 413 1985 042 0 8507260035 195256 06/16/85 DOCKET:413 CATAWBA 1 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: DUKE FACILITY OPERATOR: DUKE POWER C0.

SYMBOL: DPC ABSTRACT POWER LEVEL - 050%. ON JUNE 16, 1985, AT 1616:18:969 HOURS, A NUCLEAR INSTRUMENTATION SYSTEM HIGH FLUX RATE POWER RANGE REACTOR TRIP OCCURRED. WHILE CALIBRATING AN NIS CHANNEL AFTER PLACING IT IN A TRIPPED CONDITION, AT LEAST ONE OTHER CHANNEL TRIPPED. THE UNEXPECTED TRIPPING OF THE OTHER CHANNEL (S), SATISFYING THE 2-0VT-0F-4 LOGIC, RESULTED IN A REACTOR TRIP. THE CAUSE OF THE TRIP 0F THE OTHER CHANNEL (S) COULD NOT BE DETERMINED. UNIT 1 WAS IN MODE 1 AT 50%

REACTOR POWER AT THE TIME OF THE INCIDENT. TO REC 0VER FROM THE INCIDENT, THE POWER RANGE CHANNELS WERE RESET, AND REACTOR START-UP COMMENCED AFTER OTHER EQUIPMENT AFFECTED BY THE REACTOR TRIP WAS PLACED IN NORMAL ALIGNMENT. THIS INCIDENT IS REPORTABLE PURSUANT TO 10 CFR 50.73, SECTION (A)(2)(IV), AND 10 CFR 50.72, SECTION (B)(2)(II).

UNABLE TO LOCATE RECORD FOR LER: 413/86-025 FORM 130 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 454 1985 042 0 8505060515 194785 03/29/85 DOCKET:454 BYRON 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: SLXX FACILITY OPERATOR: COMMONWEALTH EDIS0N C0.

SYMBOL: CWE ABSTRACT POWER LEVEL - 018%. WHILE THE UNIT WAS OPERATING AT 18% POWER A REACTOR TRIP OCCURRED DUE TO POWER RANGE FLUX RATE HIGH ON THE EXCORE NEUTRON DETECTORS. INVESTIGATION OF THE SEQUENCE OF EVEN1S RECORDER SHOWED THAT THE FOUR RODS IN SHUTDOWN BANK A GROUP 1 ALL FELL CAUSING A HIGH NEGATIVE RATE. IT WAS BELIEVED THAT DEGRADATION OF CONTROL R00

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POWER FUSES AND MISSING SEISMIC CLAMPS ON THESE FUSES, AGGRAVATED BY THE SLAMMING OF A POWER CABINET DOOR WAS THE ROOT CAUSE OF THE TRIP.

THE UNIT WAS RETURNED TO POWER THE FOLLOWING DAY WITH NO FURTHER EFFECTS. HOWEVER, AN IDENTICAL TRIP OCCURRED 11 DAYS LATER AND MORE EXTENSIVE TROUBLESHOOTING WAS UNDERTAKEN. A FAULTY CIRCUIT CARD IN THE CONTROL R0D DRIVE LOGIC CABINET WAS DETERMINED TO BE THE CAUSE OF THE TWO TRIPS. THE CARD WAS REPLACED, AND THE R0D DRIVE SYSTEM IS NOW OPERATING PROPERLY.

FORM 131 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 483 1984 065 0 8501280087 192841 12/17/84 ,

DOCKET:483 CALLAWAY 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: UNION ELECTRIC CO.

SYMBOL: UEC ABSTRACT POWER LEVEL - 005%. THIS LER CONCERNS 2 EVENTS IN WHICH ESF'S WERE ACTUATED BY SG LEVEL OSCILLATIONS. ON 12-17-84 A FWIS, AFAS, AND SG BLOWDOWN ISOLATION SIGNAL WERE INITIATED BY A HI-HI LEVEL IN SG 'C.'

A FAILED CURRENT-PRESSURE CONVERTER IN THE CONTROL CIRCUIT FOR THE 'C' MAIN FEED REGULATOR VALVE WAS THE CAUSE OF THE FIRST EVENT. THE CONVERTER WAS REPLACED ON 12-17-84. CN 12-18-84 A HI-HI SG LEVEL ACTUATED A TURBINE TRIP AND FWIS. THE SUBSEQUENT DROP IN SG LEVELS THEN INITIATED A REACTOR TRIP. AS FEEDWATER CONTROL WAS BEING TRANSFERRED TO THE MAIN REGULATING VALVES IT BECAME APPARENT THE 'C' MAIN REGULATOR VALVE WAS NOT OPERATING CORRECTLY, AND SG LEVEL OSCILLATIONS INCREASED UNTIL THE AB0VE ACTUATIONS OCCURRED.

INVESTIGATION REVEALED THAT THE CONTROL VALVE WAS NOT SEATED PROPERLY.

THE VALVE WAS REPOSITIONED AND RETURNED TO SERVICE. THE GENERIC I/P CONVERTER CALIBRATION PROCEDURE WILL BE REVISED TO ENSURE THAT VALVES ARE CORRECTLY POSITIONED DURING CONVERTER CALIBRATION.

FORM 132 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 483 1985 034 0 8508210273 196138 07/18/85 DOCKET:483 CALLAWAY 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: UNION ELECTRIC CO.

SYMBOL: UEC ABSTRACT

POWER LEVEL - 100%. AT 0712 CDT ON 7/18/85, A REACTOR TRIP OCCURRED AS A RESULT OF A POWER RANGE HIGH NEGATIVE FLUX RATE SIGNAL. A FEEDWATER ISOLATION, AUXILIARY FEEDWATER ACTUATION, AND STEAM GENERATOR BLOWDOWN ISOLATION OCCURRED AS A RESULT OF THE TRIP. THE REQUIRED SAFETY-RELATED EQUIPMENT PERFORMED AS DESIGNED EXCEPT FOR ONE SOURCE RANGE AND ONE INTERMEDIATE RANGE CHANNEL OF NUCLEAR INSTRUMENTATION. THE HIGH NEGATIVE FLUX RATE OCCURRED WHEN CONTROL RODS DROPPED DUE TO A LOSS OF BOTH MOTOR / GENERATOR (M/G) SETS WHICH SUPPLY THE R0D DRIVE MECHANISMS. THE M/G SETS WERE LOST WHEN AN OVERV0LTAGE ARREST 0R FAILED DUE TO INADEQUATE COOLING OF A R0D DRIVE POWER CABINET. THE PROBLEMS WITH THE NUCLEAR INSTRUMENTATION CHANNELS WERE IDENTIFIED AS EQUIPMENT FAILURES AND THE NECESSARY REPLACEMENTS WERE MADE. THE INADEQUATE COOLING OF THE POWER CABINET RESULTED FROM THE SHUTDOWN OF THE M/G SET ROOM COOLER AND THE PRESENCE OF A TEMPORARY COOLING DUCT AB0VE THE POWER CABINETS (WHICH INHIBITIED NATURAL CIRCULATION THROUGH THE CABINET). ACTIONS TO PREVENT RECURRENCE INCLUDE UPGRADING THE COOLING CAPACITY .

FORM 133 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 483 1985 048 0 8512100060 197728 11/02/85 DOCKET:483 CALLAWAY 1 TYPE:PWR REGION: 3 t:SSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: UNION ELECTRIC CO.

SYMBOL: UEC ABSTRACT POWER LEVEL - 100%. AT 0203 CST ON 11-2-85, A REACTOR TRIP OCCURRED AS A RESULT OF A POWER RANGE HIGH NEGATIVE FLUX RATE. A FEEDWATER ISOLATION, AFW ACTUATION, SG BLOWDOWN AND SAMPLE ISOLATION, AND TURBINE TRIP OCCURRED AS A RESULT. THE REQUIRED SAFETY EQUIPMENT PERFORMED AS DESIGNED. AT 0112 DURING PERFORMANCE OF A CONTROL R0D PARTIAL MOVEMENT TEST ON R00 CONTROL BANK C, REACTOR OPERATORS RECEIVED A R0D CONTROL URGENT FAILURE ALARM AND A R0D POSITION INDICATOR DEVIATION OR POWER RANGE TILT ALARM IN THE CONTROL ROOM. AN EQUIPMENT OPERATOR DISCOVERED AN URGENT ALARM ON R0D CONTROL POWER CABINET 1AC. FOLLOWING THE CALLAWAY ANNUNCIATOR RESPONSE PROCEDURE, THE REACTOR OPERATOR RESET THE URGENT FAILURE ALARM AND THE ALARM CLEARED. WHILE ATTEMPTING TO WITHDRAW R0D CONTROL BANK C'THE RODS DROPPED TO THE BOTTOM. THE POWER RANGE HIGH NEGATIVE FLUX RATE REAC10R TRIP RESULTED WHEN A CONTROL R0D BANK DROPPED. THE BANK DROPPED DUE TO A LOGIC ERROR RESULTING FROM A FAILED CARD IN A R0D CONTROL POWER CABINET. (A SLAVE CYCLER COUNTER CARD FOR THE 1AC R0D CONTROL POWER CABINET.) THE CARD HAD FAILED AND TRIGGERED A LOGIC ERROR AND ZERO CURRENT ORDERS TO BOTH THE STATIONARY AND MOVABLE CONTROL R0D COILS. DUE TO PRIOR FAILURES OF SIMILAR CARDS, AN INVESTIGATION, WHICH INCLUDES MANUFACTURER ASSISTANCE HAS BEEN INITIATE

D. PROCEDURE

S HAVE BEEN CHANGED TO MODIFY ACTIONS TO BE TAKEN

FOLLOWING A R00 CONTROL URGENT FAILURE ALARM.

FORM 134 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 483 1985 054 0 8601300090 197706 12/26/85 DOCKET:483 CALLAWAY 1 TYPE:PWR REGION: 3 NSSS:WE ARCHITECTURAL ENGINEER: BECH FACILITY OPERATOR: UNION ELECTRIC CO.

SYMBOL: UEC ABSTRACT POWER LEVEL - 100%. ON 12-26-85 AT 1022 CST, A REACTOR TRIP OCCURRED AS A RESULT OF A SPURIOUS OVER TEMPERATURE DELTA TEMPERATURE (OTDT)

SIGNAL. AN AFW ACTUATION, FEEDWATER ISOLATION AND TURBINE TRIP OCCURRED AS A RESULT. THE REQUIRED SAFETY FEATURES RESPONDED AS DESIGNED. THE CAUSE OF THE TRIP WAS DETERMINED TO BE A SPURIOUS TRIP SIGNAL GENERATED BY LOOP 1 OTDT INSTRUMENTATION. PRIOR TO THE TRIP LOOP 3 OTDT CIRCUITRY WAS PLACED IN THE TRIPPED CONDITION TO FACILITATE TROUBLESHOOTING OF NEGATIVE SPIKES ON NUCLEAR INSTRUMENTATION CHANNEL N43. WHEN THE SPURIOUS SIGNAL OCCURRED ON LOOP 1, THE 2 OUT OF 4 SOLID STATE PROTECTION SYSTEM LOGIC NECESSARY TO INITIATE A PEACTOR TRIP WAS SATISFIED. OPERATORS TOOK IMMEDIATE ACTION TO STABILIZE THE PLANT AND INITIATED TROUBLESHOOTING OF LOOP 1 OTDT INSTRUMENTATION. AN OTDT REACTOR TRIP IS DESIGNED TO PROTECT THE CORE FROM DAMAGE DUE TO DEPARTURE FROM NUCLEATE BOILING (DNB). THE CAUSE OF THIS TRIP WAS A SPURIOUS SIGNAL, THEREFORE, AN ACTUAL DNB CONDITION WAS NOT APPROACHED.

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STEPHEN D. WILSON. REGULATORY COMPLIANCE ENGINEER BD5 5 B 5 L 17 0 5 II an . t . e e ., .. tvu .e. . i. v =, , nei C0we0%8%t " )( # fo enfs' ,,

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On September 17, 1986, with the unit in Mode 5 (Cold Shutdown) Instrumentation and i Controls (I&C) technicians were performing Surveillance Test Procedure (STP) I-57,

" Calibration of Emergency Borate Flow Channel 113." During testing, it was discovered that the reference cable between the sensing hea;1 and transmitter for channel 113 had electrical leads that were miswired, rendering the instrument

, inoperable, i

It was determined that when STP I-57 was last performed on April 1, 1985, two of l the reference cable electrical leads were incorrectly wired at the conclusion of i the procedure. On May 1, 1985, the unit entered an applicable mode of Technical

Specification (TS) 3.3.3.5 and on May 8 exceeded the time requirement for the i

action statement of TS 3.3.3.5 due to channel 113 being inoperable. On

! September 17, 1986, Channel 113 was properly wired and returned to an operable I status. During the period t_ hat channel 113 was inoperable, instrumentation for monitorint other emergency boranon flowpaths was available. An invesu gation i,s being contucted to deters'ne the root cause ano corrective actions for this event.

A supplemental report will be submitted at a later date.

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_SMOGEO99P' 861016 PDR ADOCK 05000275 M S PDR yQGf l

11395 V \

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PACIFIC OAS AND E LE C T RIC C O M PANY 3 b# '1 ,

77 BEALE STREET . S AN FRANCISCO. CALIFORNIA 94106 . (415) 781 4231

Tax 910-372-6587 saurs D.sMIPPER

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October 16, 1986 PGandE Letter No.: DCL-86-296 Document Control Desk U.S. Nuclear Regulatory Commission Nashington, D.C. 20555 Re: Docket No. 50-275, OL-DPR-80 Diablo Canyon Unit 1 Licensee Event Report 1-85-041-00 Emergency Borate Flow Channel Inoperable Due to a Hiring Error Gentlemen:

Pursuant to 10 CFR 50.73(a)(2)(1)(B), PGandE is submitting the enclosed Licensee Event Report concerning an inoperable emergency borate flow channel.

A supplemental report will be submitted at a later date.

This event has in no way affected the public's health and safety.

Kindly acknowledge receipt of this material on the enclosed copy of this letter and return it in the enclosed addressed envelope.

Sincerely, O '

1._ [w .

Enclosure cc: L. J. Chandler J. B. Martin M. N. Mendonca B. Norton H. E. Schierling CPUC Diablo Distribution INPO 1139S/0047K/JHA/1127

DC2-86-TI-N109 0

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ABSTRACT THE CHANNEL III FEEDWATER FLOW INDICATOR FOR THE "C" STEAM GENERATOR

$ BEGAN SPIKING HIGH PERIODICALLY. THE CHANNEL WAS PLACID IN A TRIPPED d _

CONDITION WITHIN ONE HOUR. THEMULTIPLIER/ DIVIDER /SQUAREROOTCARD FOR THE STEAM GENERATOR IC FEEDWATER. FLOW CHANNEL III VAILED CAUSING THE OUTPUT OF THE CARD TO PERIODICALLY, SPIKE HIGH. -THE CARD WAS '

REPLACED AND ,THE LOOP WAS CALIBRATED AND RETURNED TO SERVICE.

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\ FORM 39 LER SCSS DATA ,

07-22-86 '

DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE I

" 338 1982 074 0 8212150034 179767 11/19/82

. s DOCKET:338 NORTH ANNA 1' TYPE:PWR REGION: 2 NSSS:WE

( ,

ARCHITECTURAL ENGINEER:r3WXX .

FACILITY OPERATOR:iVIRGINIA ELECTRIC & POWER'CO. '

SYMBOL: VEP ,

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' ABSTRACT

~ ~ C ,(PROTECTION CHANNEL I OVERTEMPERATUREsDELTA~ T AND OVERPOWER DETA T WAS PLACED IN TRIP RENDERING THE CHANNEL INOPERABLE. THIS EVENT IS CONTRARY TO TECH SPEC 3.3.1.1 AND REPORTABLE PURSUANT TO TECH SPEC l 6.9.1.9.B. THIS CHANNEL WAS IN0PERABLE BECAUSE TH: G-HOT NRA CARD

': (TM-412M) FAILED. THE CARD HAD BEEN It! STALLED TWO DAYS EARLiEP, FOLLOWING A SIMILAR FAILURE. THE CHAriML WAS CALIBRATED WITH SATISFACTORY RESULTS AN3* RETURNED TO SERVICE.

~

't~ FORM ( 40 ER SCSS DATA s W 86 s

- - - - E*************************************l*******1 >.*******

DOCKET YEAR LER NUMBER' PEVISION s DCS NUMBER NSIC EVENT DATE -

338 1983 026 0 8305270323 182944 04/26/83 i i DOCKET:338 NORTH >ARMA 1 TYPE:PWP. ! '

REGION: \2 NSSS:WE ARCHITECTURAL ENGINEER: WXX FACILITY OPERATOR: VIRGINtA ELECTRIC & POWER C0. /

, SYMBOL: VEP s

ABSTRACT -

ON APRIL 16, 1983 WITH UNIT 1 AT 100 PERCENT POWER, A CALIBRATION CHECK 0F DELTA T/TAVG PROTECTION CHANNEL I REVEALED AN ERRATIC TAVG LEAD LAG DERIVATIVE CARD. 30NCONSERVATIVE CHANNEL I OVERTEMPERATURE DELTA T SETPOINTS COULD HAVE RESULTED. THE REDUNDANT CHANNELS WERE OPERABLE AND THE~AFFECTED CHANNEL WAS PLACED IN THE TRIPPED CONDITION AS REQUIRED BY THE ACTION STATEMENT OF TECH SPEC 3.3.1.1. THIS EVENT IS REPORTABLE PURSUANT TO TECH SPEC 6.9.1.9.A. THE CAUSE OF THE TAVG LEAD LAG DERIVATIVE CARD'S ERRATIC BEHAVIOR HAS NOT BEEN LETERMINED.

THE CARD WAS REPLACED, CALIBRATED, AND THE CHANNEL RETURNED T0 i

i i

SERVICE. THERE ARE N0 GENERIC IMPLICATIONS ASSOCIATED WITH THIS EVENT.

FORM 41 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 338 1985 017 0 8510250182 197326 09/17/85 i

- - - - l DOCKET:338 NORTH ANNA 1 TYPE:PWR I REGION: 2 NSSS:WE i ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: VEP ABSTRACT POWER LEVEL - 100%. ON 9-17-85 AT 1118 HRS A UNIT 1 REACTOR TRIP / TURBINE TRIP WAS MANUALLY INITIATED. THE UNIT 1 CONTROL ROOM 4 OPERATOR OBSERVED THAT THE 4 CONTROL BANK D, GROUP 1 CONTROL RODS HAD DROPPED INTO THE CORE VIA INDIVIDUAL ROD POSITION INDICATORS AND NUCLEAR INSTRUMENTATION INDICATION. THE REACTOR AND TURBINE WERE TRIPPED IN ACCORDANCE WITH THE IMMEDIATE ACTION REQUIREMENTS OF THE REACTOR TRIP OR SAFETY INJECTION PROCEDURE. ALL PLANT PARAMETERS RESPONDED NORMALLY. THE CAUSE OF THE DROPPED CONTROL RODS WAS DETERMINED TO BE AN INTERMITTENT FAULT IN THE ALARM CIRCUIT CARD t

ASSOCIATED WITH GROUP 1 0F SHUTDOWN BANK B, CONTROL BANK B AND CONTROL BANKD(POWERCABINETIBD). THERE ARE 4 SUCH ALARM CIRCUIT CARDS IN THE R0D CONTROL SYSTEM, 1 IN EACH OF THE 4 POWER CABINETS. TWO OTHER ALARM CIRCUIT CARDS WERE DETERMINED TO ALSO HAVE INTERMITTENT FAULTS AND 1 WAS PROPERLY FUNCTIONING (CABINET 1AC). THE ALARM CIRCUIT CARD IN POWER CABINET IBD WAS REPLACED WITH THE OPERABLE ALARM CIRCUIT CARD FROM POWER CABINET 1AC AS REPLACEMENT CARDS WERE NOT AVAILABLE ON l

SITE. A UNIT 1 REACTOR STARTUP COMMENCED AND CRITICALITY WAS REACHED AT 2014 ON 9-17-85, WITHOUT INCIDENT. THE 3 FAULTY ALARM CIRCUIT CARDS WERE REPLACED ON 9-18-85. THIS EVENT IS REPORTABLE PURSUANT TO 10CFR50.73(A)(2)(IV).

FORM 42 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1981 004 1 8102030421 163652 01/07/81 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: VEP ABSTRACT THE FLUX PENALTY SUMMING AMPLIFIER (NM-432A) WAS FOUND TO BE FAILED DURING A LOOP CALIBRATION CHECK. THE CAUSE WAS A FAULTY NSA CARD

(C3-141). THE CARD WAS REPLACED AND CALIBRATED.

FORM 43 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1981 012 0 6102180522 164171 01/21/81 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER CO.

SYMBOL: YEP ABSTRACT CONTAINMENT PRESSURE PROTECTION CHANNEL III FAILED LOW. THE CHANNEL WAS PLACED IN TRIP AND DECLARED INOPERABLE. THE EVENT WAS CAUSED BY

, THE FAILURE OF A POWER SUPPLY CARD TO REACTOR CONTAINMENT PRESSURE PROTECTION CHANNEL III. THE FAILED CARD WAS REPLACED WITH A CALIBRATED CARD, FUNCTIONALLY TESTED AND RESTORED TO A NORMAL CONDITION.

FORM 44 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1981 016 0 8102250636 164314 02/05/81 00CKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE ARCHITECTURAL ENGINEER: SWXX FACILITY OPERATOR: VIRGINIA ELECTRIC & POWER C0.

SYMBOL: VEP ABSTRACT THE TEMPERATURE INDICATOR (TI-412D), ASSOCIATED WITH TAVG PROTECTION FOR LOOP A, WAS READING APPR0XIMATELY 5F-15F LOW AND RESPONDING ERRATICALLY AFTER BEING PLACED IN SERVICE UPON COMPLETION OF A PERIODIC TEST ON THE CHANNEL. THE ERRATIC AND LOWER THAN NORMAL READINGS ON THE TEMPERATURE INDICATOR WERE CAUSED BY A FAILED RELAY CARD. THE CARD WAS REPLACED, CALIBRATED AND RETURNED TO SERVICE.

FORM 45 LER SCSS DATA 07-22-86 DOCKET YEAR LER NUMBER REVISION DCS NUMBER NSIC EVENT DATE 339 1981 038 0 8106090570 166494 05/05/81 DOCKET:339 NORTH ANNA 2 TYPE:PWR REGION: 2 NSSS:WE 1

ML20215E892

Contents

Text

SUMMARY

1.0 INTRODUCTION

7.0 REFERENCES

1.0 INTRODUCTION

D. PROCEDURE

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