ML20206S412
| ML20206S412 | |
| Person / Time | |
|---|---|
| Site: | Quad Cities  |
| Issue date: | 07/21/1986 |
| From: | Delgeorge L COMMONWEALTH EDISON CO. |
| To: | James Keppler NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| References | |
| 1847K, NUDOCS 8609220205 | |
| Download: ML20206S412 (19) | |
Text
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Commonwealth Edison i
72 West Adams Street, ChicaQo, Illinois V
Address Reply to: Post Office Box 767 Chicago, Illinois 60690-0767.
July 21, 1986 Mr. James G. Keppler Regional Administrator U.S. Nuclear Regulatory Commission Region III 799 Roosevelt Road Glen Ellyn, IL 60137 subject: Quad Cities Station Units 1 and 2 l
Response to IE Inspection Report i
Nos. 50-254/86002 and 50-265/86002 NRC Docket Nos. 50-254 and 50-265
Reference:
-(a): Letter from C. E. Norelius to Cordell Reed dated May 30, 1986.
Dear Mr. Keppler:
This letter is in response to the inspection conducted by Messrs.
A. L. Madison and A. D. Morrongiello of your staff on February 9 through April 12, 1986 of certain activities at Quad Cities Station. The referenced letter indicated that certain activities appeared to be in noncompliance with NRC requirements. The Commonwealth Edison Company's response to the Notice of Violation is provided in Attachment A.
The referenced letter also requested we describe actions we are taking to address a recent increase in personnel errors and the adequacy of 50.59 reviews. Our response to these concerns is provided in Attachment B.
This response is provided consistent with an extended due date granted during telecons with Duane Boyd (June 23) and Bill Guldemond (July 14).
't If you have any further questions regarding this matter, p'aase contact this office.
Very truly yours, 8609220205 860721 PDR ADOCK 05000254 e
L. O. i D$1 George Assistant Vice-President la Attachment cc: NRC Resident Inspector - Quad Cities ito /
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4 ATTACMENT A CON 00NWEALTH EDISON COMPANY RESPONSE TO NOTICE OF VIOLATION As a result of the inspection conducted on February 9 through April 12, 1986, the following violations were identified:
ITEM OF NONCOMPLIANCE 1.
10 CFR 50.59 requires that adequate design reviews be performed when changes are made to the facility that affect the safety analysis report.
Contrary to the above, several facility modifications were accomplished, including the addition of Anticipated Transient Without Scram (ATWS)
Diverters, which added electrical load to the station 125V batteries without determining the effect of this additional load on the batteries. This resulted in the batteries and the battery chargers being found on May 11, 1984 incapable of sustaining design full load discharge for the prescribed time requirement in the safety analysis report.
DISCUSSION Past modifications which added a new DC load (generally a fraction of a percent of total battery capacity) used engineering judgement to determine that the additional load was acceptable. Engineering judgement was the only vehicle available to the reviewer, as a detailed load study was not available, and the FSAR (Section 8) did not provide detailed load figures for 125 VDC loading. The general conclusion for any one particular modification being reviewed was that the small additional load relative to battery size would only consume a small part of the battery design margin; therefore, the added load would not jeopardize battery performance.
The concern of exceeding battery capacity cannot be attributed to any one particular modification, including the ATWS inverter modification but rather to the cumulative effect of a number of small loads added over an extended period of time.
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J CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED The situation identified in the noncompliance where the 125V batteries and chargers were found to be inadequate was addressed by a Confirmatory Action Letter (CAL) dated May 7, 1984. Our response to the CAL dated May 11, 1984 documented our actions in response to this concern.
Specifically, a 125V DC load shedding procedure was implemented (May 5, 1984) to shed non-essential DC loads. Our response included a battery profile analysis which demonstrated sufficient capacity to accomodate both accident and non-accident scenarios (crediting the load shed Procedure).
In addition to the above, a more detailed 125V DC load profile study was completed on November 2, 1984. The results of this study confirmed that the 125V DC batteries could supply all essential DC loads. Also, as noted in the Inspection Report, the 125V batteries and chargers at Quad Cities have been replaced with higher capacity units.
With respect to the overall concern of ensuring that modification design reviews address battery capability, corrective action was actually taken prior to the May 1984 sequence of events. On January 18, 1982, the Systems Interaction Checklist was added to Station Nuclear Engineering Department (CNED) Procedure Q.6 which is the procedure used to establish the control and processing of modifications. The checklist questions if additienal battery loading has been considered. All modifications receiving approval after this date would contain this checklist as supporting documentation that battery loading was considered.
Modifications approved prior to this date, such as the ATWS inverter modification (approved July 24, 1979) would lack this documentation.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED l
Full compliar.ce was achieved on November 2, 1984 when the detailed load profile study was completed. The SNED Systems Interaction Checklist, the load shedding procedure and the load profile study together ensure that the battery capability meets FSAR requirements and that future l
modifications will be reviewed with respect to additional battery j
loading.
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.. ITEM OF NONCOMPLIANCE 2.
10 CFR 50.72(b)(2)(ii) requires that any event that results in manual or automatic activiation of any Engineered Safety Feature, including the Reactor Protection System (RPS), be reported to NRC within four hours, unless said activation resulted from and was part of a preplanned sequence during testing or operation.
Contrary to the above, while in a Refueling Outage, the following events occurred on Unit 1:
a.
On March 10, 1986, an ATWS/ Alternate Rod Insertion (ARI) activation occurred. The cause of the event was a valving error and not as part of a preplanned sequence.
b.
On March 17, 1986, a spurious ATWS/ARI was received due to contractor personnel bumping a sensitive line.
In both cases, the required four hour phone call was not made and notification to the NRC was not planned until the resident inspectors notified.the licensee of the requirement to do so.
This is considered a repeat violation (See Inspection Report No. 254/85002).
2.
DISCUSSION a) LER 254/86-013 describes the details of this event. Briefly, an error was made during the performance of QIS-47, Function Test of the Excess Flow Check Valves. Unit One was in the REFUEL mode and in the process of performing the 1000 psig Primary Systems Leakage Test. The result of the error was an ATWS trip. The operating department shift personnel did not make a four hour ENS notification because the Instrument Mechanic had warned the operator that there was a potential for a scram, recirculation pump trip, and ECCS initiation. They felt this satisfied the
" preplanned sequence of events" criteria.
b.
LER 254/86-019 describes the details of the second event. Briefly, contractors working on scaffolding adjacent to the ATWS level instrument sensing lines momentarily disturbed the lines causing an initiation signal to exist for several seconds. Several SRO licensed personnel debated if the ATWS system was an Engineered
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Safety Feature (ESP) System. They referred to NUREG-1022 Supplement 1 for guidance. Question 6.1 on page 7 of the NUREG j
says the criterion is based on ESF systems being defined in the FSAR. The ATWS' system is not in the original FSAR nor described as
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an Engineered Safety Feature in the updated FSAR. The system is
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not considered part of the Reactor Protection System (RPS).
Consequently a decision was made that the event was not reportable.
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.. CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED a.
Orginally this event was classified as non-reportable, but was re-classified on March 11, 1986, due to the fact that the ATWS signal received during step F.18 was not preplanned since it was the result of an error. A four hour Emergency Notification System (ENS) notification was made in accordance with 10 CPR 50.72 on March 11, 1986, at 1240 hours0.0144 days <br />0.344 hours <br />0.00205 weeks <br />4.7182e-4 months <br />.
b.
A discussion was held with the resident inspector on March 24, 1986 and the reportability of the ATWS system was clarified.
It was mutually agreed to in this meeting that ATWS will be considered as an ESF system. The DVR was reclassified as an LER on this day.
i Furthermore, in response to this non-compliance and the concern regarding repeat violations, a discussion was held with the resident inspector on July 2, 1986 to clarify when an ENS notification is or is not required. The results of this discussion were relayed to the operating department personnel during the weekly meetings held on July 3, 1986 and July 10, 1986. The previous determination that the ATWS system would be considered as an ESF system and the proper interpreta-tion of " preplanned sequence of events" was also discussed in these meetings.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved after the weekly meeting of July 10, 1986.
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.. ITEM OF NONCOMPLIANCE 3.
Technical Specifications Section 6.2 requires that detailed written procedures, including applicable checkoff lists shall be prepared, approved and adhered to for preventive and corrective maintenance which could have an effect on the safety of the facility; normal startup, operation, and shutdown of the reactor and systems and components involving nuclear safety of the facility;.and Surveillance and Testing requirements.
Contrary to the above, the following are examples of failure to adhere to procedure:
a.
On January 2, 1986, an Equipment Operator was directed to place the IB reactor building fan out of service pursuant to a master out of service checklist for work request No. 46726. The operator mistakenly removed the fuses for the 1/2 swing Emergency Diesel Generator (EDG) Cooling water pump making the EDG inoperable.
b.
On January 22, 1986, while troubleshooting the Unit I refuel floor radiation monitor, Channel B, for work request No. 47147, the amphenol for the reactor building ventilation detectors was pulled in error, causing isolation of the reactor building ventilation.
c.
On February 13, 1986, while attempting to couple control rods after maintenance on Unit 1, the Unit Operator incorrectly placed the Mode Selector switch to "Startup" and received a full scram. The Scram Discharge Volume (SDV) was full because of having its vent and drain valves out of service and valve leakage into the SDV. Placing the mode selector switch to startup unbypassed the SDV high level scram. Had the operator been aware of plant conditions as required by QAP 300-2 " Conduct of Operations," he would have foreseen this event.-
d.
On March 10, 1986, while performing QIS-47 " Excess Flow Check Valve I
Surveillance," Valve C7/X-29E, Anticipated Transient Without Scram (ATWS) level instrument drain valve, was opened instead of Valve C5/X-49B, reactor pressure tap drain valve. This caused an ATWS-ARI
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actuation, e.
On March 21, 1986, while performing QTS 150-6 " Integrated Primary Containment Leak Rate Test (IPCLRT)," Valve 1-1001-23A was opened instead of Valve 1-1001-26A (These are drywell spray isolation valves). This resulted in contaminated water released outside secondary containment.
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DISCUSSION On January 2, 1986, an Equipment Operator was directed to remove the 1B Reactor Building Supply Fan from service. The out-of-service checklist required removal of the Bus 18 control power transformer. fuses for the IB i -
Reactor Building Supply Fan feed breaker. Approximately two and one-half hours after the out-of-service had been completed, it was discovered that 4
the-operator had mistakenly removed the fuses for the 1/2 Diesel Generator Cooling Water Pump instead of the intended fuses. Upon investigation, it was discovered that the fuse labels for Buses 18, 19, 28, and 29 were not labeled consistently, which resulted in operator confusion. The labels for the fuses at Bus 28 and 29 are located directly above their corresponding fuses. At Buses 18 and 19, however, the labels are located below their corresponding fuses, and, in many instances, are closer to the fuses below them than to the correct fuses above them.
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CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED Corrective action regarding the incorrect out of service of the shared 1/2 emergency diesel generator has been completed. To provide a better system design of the control power transformer fuses additional labeling has been added to the engraved name plates. This consists of placing arrows on the labels for all fuses in Busses 18, 19, 28 and 29 pointing to the corresponding fuses to prevent further confusion. The operator involved with this event was counseled by station management on the significance of the error and the importance of making sure that the correct equipment is worked on.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATIONS The corrective action taken to avoid a recurrence of this type was to expand the out of service verification program. Prior to this event, the out-of-service verification program required verification of safety related equipment only. Since the fan to be taken out-of-service in this event was non-safety related, the work was not verified. Independent verification of this out-of-service, which was on a safety-related bus, would have detected the error while the job was still in progress. The j
verification program now includes a second independent verification of all out of services on safety related busses plus any non-safety related out of service which is determined to have a possibility of affecting safety related equipment.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The labeling of the control power transformer fuses was completed on January 2, 1986. The second verification of out of services on safety related busses was also instituted on January 2, 1986.
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3.b DISCUSSION The Technical Staff and Instrument Maintenance personnel involved in this event were in the process of performing a field verification of wiring for the IB Fuel Pool Monitor. This process was specifically performed during unit shutdown so that possible safety consequences would be minimal. The verification required that the J4 amphenol connector of the IB Fuel Pool Radiation Monitor Trip Unit be disconnected. Inadvertently, the J4 amphenol connector for the IB Reactor Building Ventilation Radiation Monitor on the main chassis was disconnected and initiated this event.
Personnel unfamiliarity with the equipment involved is the root cause of this event.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED Discussions of this occurrence and its significance were held immediately after the event on January 22, 1986 with the personnel involved. Station
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management counseled them on the importance of verification that the correct equipment, is being worked on prior to performing the work. The personnel understood the implications of this event.
1 CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATIONS Additionally, on April 16, 1986, discussions of this occurrence, and other occurrences of a similar nature, were held with the station department heads to be shared at departmental tailgate sessions, stressing the implications of the event and the importance of an absolute understanding of all facets of the task to be performed. Included in this understanding should be what task is to be performed, how to perform the task, equipment th> task is to be performed on and most importantly, if any doubt exists on any aspect of the task, questions should be asked to remove that doubt.
This occurrence and others of a similar nature will be reviewed again in a department heads meeting prior to the next refueling outage to be discussed at weekly station department meetings. Emphasis will be placed on preventing errors of a similar nature during refueling outage i
activities.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The station is currently in full compliance.
! 3.c CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED The corrective action taken at the time of the occurrence was to place the Mode switch in Refuel and reset the scram signal. Proper notifications were completed as required.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATIONS To prevent recurrence of an incident of a similar nature tailgate sessions were held with operations personnel. During these sessions, this incident and other errors were discussed. Everyone was reminded that it is important to pay attention to detail, ask questions, and not to get in a hurry in order to prevent errors. This incident will also be discussed at a department heads meeting prior to the next refueling outage with the information to be shared at weekly station department meetings. Emphasis will be placed on preventing errors of a similar nature during refueling outage activities.
2 DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Tailgate sessions were held on February 27, April 17, and May 15, 1986.
3.d CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED The cause of the valving error during the performance of QIS-47, " Excess Flow Check Valve Surveillance," was inadequate personnel communications.
The corrective action taken was to discuss this event with all Instrument Maintenance personnel on March 20, 1986 during a weekly department meeting to make them aware of the cause. The requirement for accurate communication during surveillance activities was stressed.
i CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATIONS The instrument rack drain valves will be inspected to verify proper valve identification tags are in place. Missing tags will be replaced.
In addition, the station has ordered specially colored tags which will be attached to the valves that are used specifically for this procedure.
Placement of these tags will aid in identification of the valves used for the procedure.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The placement of the noriaal valve identification tags and the special identification tags for both Unit 1 and Unit 2 will be completed prior to October 15, 1986, the beginning of the Unit 2 refueling outage (QIS-47 is a refuel outage surveillance).
.. 3.e DISCUSSION
'On March 21, 1986 Quad Cities Unit One was being prepared for performance of the Integrated Primary Containment Leak Rate Test (IPCLRT). By 2300 houra all that remained was to pressurize the drywell containment. This was to be accomplished by the use of an auxiliary air compressor located outside the reactor building. The discharge of the external compressor was connected to the Unit One drywell containment via hose to a fire system piping penetration of the reactor building, through manual isolation valves,'through specially installed piping to a flaage on the RHR containment spray piping which connects with the drywell by way of a containment spray header. The exact point of connection to the containment spray flange was between the 1-1001-23A and 1-1001-26A motor operated valves, which are the injection valves for the containment spray function of the "A" loop of the Residual Heat Removal (RHR). system.
At 2300 hours0.0266 days <br />0.639 hours <br />0.0038 weeks <br />8.7515e-4 months <br />, the air compressor had been started and the final fire system penetration isolation valve had been opened. This provided air pressure to the volume between the 23A and 26A valves. The test director then instructed the Unit One NSO to open the 1-1001-23A valve. By opening the 1-1001-23A valve, the auxiliary air compressor discharge was connected - not to the drywell containment spray header -
but to the pressurized RHR system which was being used for reactor shutdown cooling at the time. The result was reactor water traveling back to the compressor and a small amount of leakage on the ground.
The test director should have specified opening the 1-1001-26A valve to begin containment pressurization.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED The test director involved in this event is fully aware of the j
significance of his error and the serious nature of the role of test director. This event and his performance were discussed in a meeting with the Assistant Superintendent of Technical Services. The weaknesses of the test method and performance were discussed and the test director himself identified areas where his performance was inadequate.
CORRECTIVE ACTION TO AVOID FURTHER VIOLATIONS Procedures QTS 150-1 and QTS 150-6 will be revised to include a more rigorous description of how to accomplish drywell containment pressurization. Specific reference will be made to sequence of events and valve numbers. The procedure revisions will include requiring the 1001-23A valve be taken out of service during the performance of this test.
.. A report of this event was distributed to all department heads so that it could be discussed at individual station department meetings. Special emphasis was placed on discussing the philosophy of double-checking and preventing errors from becoming operational problems through teamwork.
A check valve will be added to the fire system penetration piping on the inside of the reactor building. This check valve will be utilized in all future containment leak rate tests. Addition of this valve will provide additional protection against the backflow of any fluid outside the Reactor Building, thus preventing contaminated water from being released outside secondary containment.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The procedure revisions and the addition of a check valve will be completed prior to the performance of the next containment leak rate test.
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.. ITEM OF NONCOMPLIANCE 4.
10 CPR 50.59 requires that adequate design reviews be performed when changes are made to the facility.
Contrary to the above, the following are examples of inadequate design reviews that led to unnecessary safety system challenges and actuations:
a.-
On February 11, 1986, a full reactor scram was received on Unit 1 from SDV high level. As a part of the HFA relay modifications, the SDV vent and drain valves were closed. Water accumulated in the SDV due to valve leakage. The SDV high level scram was bypassed with its keyswitch,.however, removal of the 111 relay for the HFA modification disabled the bypass. This should have been foreseen as a part of the design review.
b.
On February 11, 1986, Group II and III isolations were received on i-Unit I while removing a telay for HPA modifications. It was later determined that this relay was powered in series with several other relays including the reactor low water level relay which initiated the isolations, thus removing one relay de-energized several relays. This should have been foreseen as part of the design review.
c.
On February 21, 1986, while instrument maintenance personnel were installing new main steam line (MSL) radiation monitors on Unit 1, the reactor building ventilation system isolated and standby gas treatment auto-started.
It was later determined-that MSL radiation monitors were powered in series with the ventilation and l
fuel pool radiation monitors, thus, disconnecting one de-energized j
all and resulted in the above actuations. This should have been l
foreseen as part of the modification design review.
1 d.
On March 3, 1986, it was found that a modification had been accomplished under work request No. 45346 on Unit 2 which inadvertently routed High Pressure Coolant Inspection (HPCI) gland exhauster tube side water to the shell side, flooding the Turbine, making HPCI inoperable, and leading to an Unusual Event. The work l
should have been accomplished under control of a modification, however, inadequate design review resulted in the work being carried out under a work request and the work request directing the faulty correction.
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- 4.a DISCUSSION Preplanning for the HFA coil replacement modification in the 901-15 panel dictated that Unit One he in cold shutdown with no fuel in the vessel prior to the initiatLm of the relay replacement. This action was intended to preclude any events detrimental to the safe operation of the plant from occurring. As required, the 901-15 panel HFA relays were removed from service via removal of fuses to the relay circuit.
Removal of these fuses disabled the Scram Discharge Volume (SDV) high level scram bypass relays. As part of the 901-15 panel work, the SDV vent and drain valves were out-of-service in the CLOSED position, thus isolating the volume. Subsequently, a scram occurred from SDV high level due to reactor water leakage by the control rod drive scram outlet valves.
Preplanning did not anticipate the control rod drive scram outlet valve leakage nor the total amount of that leakage. Preplanning had the reactor in the safest mode available. Thus, there were no safety concerns for this event.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED The scram signal was left inserted while this portion of the HPA relay work was completed to prevent additional scrams. Proper notifications were completed as required. The control rod drive scram valves were checked for leakage as Unit One was started following the refueling outage. Four valves were identified to be leaking and repairs have i
been made to those valves.
CORRECTIVE ACTION TO BE TAKEN TO AVOID FURTHER VIOLATIONS To prevent a recurrence of this type during the Unit Two HPA relay replacement, the reactor vessel again will be in cold shutdown with no fuel in the vessel. When the Scram Discharge Volume (SDV) high level scram bypass relays are to be replaced, the relays will be individually removed from service, (independent of the 902-15 and 902-17 panel out-of-service) while the SDV vent and drain valves are in service and in *he OPEN position. This will allow water accumulation (if any) from control Rod Drive scram outlet valves to drain thus preventing a scram due to SDV high level. The modification to the bypass relays will be completed prior to or following the out-of-service for the 902-15 panel and 902-17 panel. This will provide additional assurance that the SDV vent and drain valves will not be taken out-of-service during the bypass relay replacement.
DATE WHEN PULL COMPLIANCE WILL BE ACHIEVED The station is currently in full compliance.
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DISCUSSION b.
Replacement of the "B" channel scram reset relays HFA coils was in progress when a " daisy-chain" wire configuration was inadvertently broken. This caused a trip of the Group II and Group III isolation logic and equipment.
To minimize the safety consequences of performing a modification of this type, where many wire connections were to be lifted and relanded, this work was performed during a scheduled refuel outage. Further installation review required that the modification to these relays be performed with no fuel in the reactor vessel. Thus, no safety concerns were present during this event.
c.
While attempting to connect power to the new Main. Steam Line Radiation Monitors, Instrument Maintenance personnel inadvertently broke a " daisy-chain" and disconnected the power supply to the IB 1
Puel Pool Radiation Monitor and IB Reactor Building Ventilation Monitor. This caused an unplanned Standby Gas Treatment System initiation and Reactor Building Ventilation isolation.
l CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED These events were discussed at the March 5, 1986 Department Heads Meeting. It was requested that each Department Head discuss this common practice of field wiring with their departments so that everyone, especially those who prepare work packages, is aware of the importance of checking the electrical schematic and wiring diagrams for the " daisy-chain" wiring configuration.
CORRECTIVE ACTION TO BE TAKEN TO AVOID FURTHER VIOLATIONS b.
The work packages for future HFA relay replacements will be written such that when working on relays in one electrical division panel, no work on the relays in the other electrical division panel will be allowed.
b&c Additionally, to aid in the identification of possible
" daisy-chains" during electrical work package preparation, Nuclear Work Request procedure, QAP 1500-2, will be revised to indicate that both the affected electrical schematic and wiring diagram should be reviewed for possible " daisy-chain" wiring configurations, and if necessary, jumpers / blocks will be installed to prevent inadvertent " daisy-chain" breaks. The affected electrical schematic and wiring diagrams should also accompany the work packages and be used as references during installation.
DATE WHEN PULL COMPLIANCE WILL BE ACHIEVED This procedure change will be completed by August 15, 1986.
.. 4.d DISCUSSION On October 18, 1985, the unit Operating Engineer wrote work request Q4534~1 to "...-route the (HPCI gland seal) exhauster drain to the 3/4 inch PT vent on the (HPCI) Graham condenser" with the intent of replacing the existing drain line. The existing drain was not allowing water to flow due to excessive internal corrosion. The 3/4 inch PT vent, in actuality, was on the tube side of the condenser and not the shell side where the drain was originally routed.
l The physical rerouting of the drain line was not considered to be a modification since the equipment involved (exhauster and drain line) is non-safety related and has no special design criteria. Additionally, the drain line was intended to be routed to the shell side of the condenser as had been the original design.
This event indicates a failure of the personnel involved to perform an independent verification of the recommendation of another department.
The independent verification / double checking concept is utilized extensively throughout the station such as during jumper / block installations, equipment installation testing (performed by personnel from other than the installing department), safety-related equipment out-of-services and return-to-services, on-site review of procedures, test, etc. This should have been applied in this event.
CORRECTIVE ACTION TAKEN AND RESULTS ACHIEVED To return the HPCI system to an operational status, immediately following the event, the gland seal exhauster drain line was temporarily routed to the HPCI room floor drain sump via a drain hose.
Additionally, to immediately ensure that the independent verification
/ double checking process was implemented for this installation, it was reclassified as a modification (M4-2-86-18).
The modification was successfrily completed June 5, 1986 during a weekend outage.
CORRECTIVE ACTION TAKEN TO AVOID FURTHER VIOLATIONS This event was discussed during the March 24, 1986 Department Heads meeting and at length with all Technical Staff personnel during the bi-monthly Technical Staff meetings April 16, 1986 and May 8, 1986.
Specific emphasis was placed on the engineer's responsibility as an independent verifier of design as well as emphasis on field verification of equipment / connections during the design review process.
.. To prevent recurrence of this type of an event, a checklist for development of work request projects has been implemented. This checklist indicates the documentation required to support the work request project installation. A safety evaluation in the 10 CFR 50.59 format and design review are required without exception.
With respect to the issue of whether this work constituted a modifica-tion, the job was ultimately reclassified as a modification as previously stated. To provide additional detailed guidance on determining proper classification, a revision to Ceco Quality Procedure 3-51 is being implemented to provide better definition of what constitutes a modifica-tion and what type of work can be performed under other administrative control systems. This revision is a result of the modification task force review conducted in response to the Dresden Unit 3 modification installation inspection.
Implementation of this revision is expected by September 1, 1986.
DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED The station is currently in full compliance.
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ATTACISBNT B RESPONSB TO ITEMS OF CONCERN ITEM OF CONCERN - DESIGN REVIENS The design review involves the development of the proposed modification. The design review process performed for modifications to nuclear plant systems or equipment may be divided into two areas of review, engineering and safety.
The engineering review considers the technical feasibility of the proposed modification, i.e., whether or not the modification is acceptable based on sound engineering principles.
The safety review, in the 10 CPR 50.59 format, discusses the safety implications of performing the modification, ensuring that the continued safe operation of the plant will not be adversely affected upon completion of the modification installation.
The physical installation and associated aspects of the modification are not considered to be part of the design review process.
Rather, it is considered to be part of the installation review process.
The plant status required, (i.e. outage, non-outage, load reduction, etc.), for installation of the modification is determined initially by the Operating Engineer based upon the effect of the modification on plant equipment / systems. Modification installations to be performed during outages are discussed during pre-outage planning meetings and the daily outage status meetings. Discussions include the equipment outage requirements, plant status requirements, and most importantly safety l
requirements for the installation.
It is at this point that a final determination of the plant equipment status for modification installation is identified, f
The physical installation of the safety related modifications are controlled by approved station procedures and step-by-step work instructions which have been on-site reviewed.
Post modification installation testing requires a visual t
verification and/or functional verification depending upon the type of
. installation. The test verifies that the modification was installed as 4
designed and performs as designed. The testing instructions are reviewed i
and are written to test all systems / equipment affected by the modification to assure that all systems are fully operational as required. Modifications that required a special plant operating status, (i.e., no fuel in the vessel), for installation are fully tested to assure the operational ability of the affected systems / equipment prior to removing the special plant operating status requirement.
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..- Accordingly, the station feels that the events described in 4a.,
b., and c. do not represent inadequate design reviews since installation method is not a part of the design review process. Neither do these events represent a safety concern of any type since during the. installation review process it had been determined that the modifications should be installed with no fuel in the reactor vessel.
These events demonstrate the need for additional attention to detail from an operational and procedural stand point.
Event 4a. required more attention to detail during the installation review process with regard to the operational status of equipment associated with the Scram Discharge Volume (SDV). The SDV vent and drain valves had been removed from service in the CLOSED position prior to installation of the HFA relay modification in the 901-15 panel. Thus, the valves were not available for draining water accumulated from the leaking Control Rod Drive (CRD) scram outlet valves.
Events 4b. and c. require more attention to detail during the installation review process. The preparers of formal work instructions for electrical packages need to keep in mind at all times that " daisy chain" wiring does exist in the plant. Both the affected electrical schematic and wiring diagrams should be reviewed during package preparation to identify the " daisy' chains" and ensure that they are not inadvertently broken during modification / maintenance installations.
Event 4d. was a failure of the double checking / independent verifier concept during the work request project. The Technical Staff engineer had based the design review on the recommendation of operating personnel without adequately verifying the information received. The i
Technical Staff engineer had felt there was no reason to question the recommendation since the operating personnel had several more years of experience at the station.
It is felt that doubling checking / independent verification is the obligation of all personnel at the station irregardless of the experience level of the individual. This type of event exactly represents the reason why the independent verification is a must and should be applied to all situations. The corrective actions implemented in response to Violation No.
4 provide additional procedural controls in this area as well as communication with the affected personnel to assure understanding of their responsibilities.
4 1847K
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.. ITEM OF CONCERN - PERSONNEL ERRORS Quad cities Station is dedicated to the reduction and elimination of personnel errors. In March of 1986, the Station had independently identified the increase in personnel errors beins experienced at the Station during the ongoing Unit one refueling outage (ideritified in this inspection report).
Meetings were held with upper Station management personnel to discuss and seek cures for the worsening trend. There were several conclusions as a result of these meetings. First, the activities associated and performed as part of a refueling outage involve an increased potential for errors due to the nature of the activities. Second, the reportability of ESF actuations has focused attention on the already sensitive issue of personnel errors. Third, some of the errors made were errors of lack of attention-to-detail, indicating a need to continually emphasize the importance of error free operation to all' station personnel. Fourth, some of the errors indicated a lack of station personnel double-checking each other and questioning each other on their activities.
This seemed to be an important practice to reinforce in all station personnel.
As a consequence of the station management meetings, department meetings were held to discuss the importance of eliminating errors as referenced in the LER's and DVR's written to document each event. Emphasis was placed on working together to eliminate ercors by checking on each other's activities, thereby preventing individual errors from getting through the system and becoming operational problems. The importance of individual's attention-to-detail was also stressed. The evaluation of each event with respect to the administration of disciplinary action is a continuing option that the station also intends to use when appropriate.
The performance indicator in this area that the station uses to evaluate trends, is the number of personnel errors as specified in causes for DVR's or LER's.
In the first three months of 1986, the station had 19 LER's/DVR's caused by personnel error. This time period coincided with the Unit one refueling outage and the time period covered by this inspection report.
In April and May of 1986, immediately after the outage, there were only 3 LER's/DVR's total caused by personnel error. The station will continue to monitor this performance indicator on a monthly basis with special attention paid during the next refueling outage to continue to improve in the area of personnel errors.
18474
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