IR 05000456/1989009
| ML20247A758 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 05/15/1989 |
| From: | Hinds J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III) |
| To: | |
| Shared Package | |
| ML20247A747 | List: |
| References | |
| TASK-1.G.1, TASK-TM 50-456-89-09, 50-456-89-9, 50-457-89-09, 50-457-89-9, NUDOCS 8905230269 | |
| Download: ML20247A758 (13) | |
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U.S. NUCLEAR REGULATORY COMMISSION
REGION III
Report Nos. 50-456/89009(DRP);50-457/89009(DRP)
Docket Nos. 50-456; 50-457 License Nos. NPF-72; NPF-77 Licensee: Commonwealth Edison Company Post Office Box 767 Chicago, IL 60690 Facility Name: Braidwood Station, Units 1 and 2 Inspection At: Braidwood Site, Braidwood, Illir.ois Inspection Conducted: March 19 through April 29, 1989 Inspectors:
T. M. Tongue T. E. Taylor G. A. VanSickle MAY I s 198g Approved B :
J.
. Hinds, J C
f actor Projects ection 1A Date Inspection Summary Inspection from March 19 through April 29, 1989 (Report Nos. 50-456/89009(DRP);
30-457/89009(DRP))
Areas Inspected:
Routine, unannounced safety inspection by the resident inspectors of licensee action on previously identified item; licensee event report review; regional request; follow-up on TMI action items; Unit 2 boron dilution prevention system (BDPS) inoperability; inadvertent safety injection (SI); contaminated equipment; operational safety verification; monthly maintenance observation; monthly surveillance observation; resumption of normal operations after strike by security guards; systematic appraisal of licensee performance (SALP) meeting; augmented inspection team (AIT); training effectiveness; and report review, Results: Of the sixteen areas inspected, no violations were identified in fourteen.
In the remaining areas two violations were identified regarding Unit 2 BDPS (paragraph 6) and inadvertent SI (paragraph 7).
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ADOCK 05000456 PDC E
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DETAILS 1.
Persons Contacted Commonwealth Edison' Company (CECO)
T. J. Maiman, Vice President, PWR Operations
- E. L. Martin, Director, Quality Assurance (Engineering / Construction)
- R. E. Querio, Station Manager
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- D. E. O'Brien, Technical Superintendent
- K. L. Kofron, Production Superintendent S. C. Hunsader, Nuclear Licensing Administrator G. R. Masters, Assistant Superintendent - Operations
- G. E. Groth, Braidwood Project Manager, PWR Projects Department
- R. J. Legner, Services Director M. Lohman, Assistant Superintendent - Maintenance P. Smith, Operating Engineer - Unit 1
- R. J. Ungeran, Operating Engineer R. Yungk, Operating Engineer - Unit 2 B. McCue, Operating Engineer - Unit 0
- R. D. Kyrouac, Quality Assurance Supervisor
- D. E. Cooper, Regulatory Assurance Supervisor R. Lemke, Technical Staff Supervisor A. D' Antonio, Quality Control Supervisor D. Ambler, Health Physics Supervisor E. Roche, Health Physics Group Leader
- F. D. Willaford, Security Administrator R. Byers, Site Superintendent - Work Planning and Startup W. McGee, Training Supervisor
- L. W.-Raney, Nuclear Safety Supervisor S. Hedden, Master, Instrument Maintenance R. Hoffman, Master, Mechanical Maintenance J. Smith, Master, Electrical Maintenance
- E. W. Carroll, Regulatory Assurance
- P. G. Holland, Regulatory Assurance
- H. D. Pontious, Operations Staff
- P. Smith, Unit 1 Operating Engineer
- N. Keetschmer, MIS Supervisor
- T. M. Bandura, Quality Assurance
- S. Jensen, Shift Foreman
- Denotes those attending the exit interview conducted on April 27, 1989, and at other times throughout the inspection period.
The inspectors also talked with and interviewed several other licensee employees, including members of the technical and engineering staffs, reactor and auxiliary operators, shift engineers and foremen, and electrical, mechanical and instrument maintenance personnel, and contract security personnel.
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2.
Licensee Action on Previously Identified Item-Open Item (Closed) 456/87029-02:
Integrity of Systems Outside Containment Likely to Contain Radioactive Material.
In Section 9.3.5.2 of NUREG-0876, Supplement 5, the staff's review of this item found that the program presented _by the licensee met the requirements of III.D.1.1, except that the initial leak test data had not been provided.
Subsequent to the staff's. review, the licensee submitted the leak test data letters addressed to Dr.- Thomas E. Murley, Director, Office of Nuclear Reactor Regulation- (NRR), dated September 25, 1988 and December 9, 1988. Section 9.3.4.2 of NUREG-0876 states that the licensee's program would conform to III.D.1.1 upon submittal of the leak test data. Therefore, this item is considered closed for Units 1 and 2.
No. violations-or deviations were identified.
3.
Licensee Event Report (LER) Review (92702)
Through direct observations, discussions with licensee personnel, and review of records, the following event reports were reviewed to determine that deportability requirements were fulfilled, that immediate corrective action was accomplished, and that corrective action to prevent recurrence had been or would be accomplished in accordance with Technical Specifications (TS):
(Closed)'457/89001-LL: Loss of 2B Residual Heat Removal (RHR) Loop as a Result of Procedural Deficiency. On February 23, 1989, both trains of the solid state protection system (SSPS) were in test for performance of an engineered safeguards surveillance. With the Unit in Mode 5 the 2B RHR pump was operating in the recirculation mode while the 2A RHR pump was out of service for maintenance. When a pressure signal in excess of 1448 psig was applied to one of the reactor coolant system (RCS) pressure instrumentation loops, the 2B RHR pump suction isolation valve closed. The valve automatically closes when reactor coolant system pressure exceeds 662 psig.
Upon loss of pump suction, the pump was immediately manually tripped.
The operators acted quickly to restore the RHR loop within 30 minutes, during which the RCS temperature increased only to 146 F.
Alternate methods of decay heat removal were available throughout the event. All appropriate event classifications and notifications were made in a timely manner. The event was caused by a deficient procedure, which failed to identify that putting the SSPS in test would not block the automatic closure of the RHR pump E
suction valves.
As corrective action, the licensee is revising the procedure to eliminate this deficiency.
Also, the licensee is instituting a more detailed and comprehensive review process for complex procedures involving engineered safety features and the SSPS, as well as other important procedures.
l The licensee is also conducting a detailed review of the SSPS to identify whether unexpected results like that encountered during this event would occur during the performance of other procedures. Based on these actions, the inspector has no further concerns. This LER is considered closed.
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(Closed) 456/89001-L1:
Reactor Trip Due to Spurious Loss of Output Voltage on Instrument Inverter 112. This LER was closed in Inspection Report 456/89005;-457/89005.
It has been revised solely for an administrative change to the report's sequential number. This LER is considered closed.
(Closed) 456/89003-LL:
Containment Ventilation, Fuel Handling Ventilation, Control Room Ventilation Actuations Due to Momentary Loss of Voltage to Their Associated Radiation Monitors. On March 2, 1989, a voltage perturbation on the 345 KV transmission system resulted in several area and process radiation monitors reverting to their interlock positions. As a result, containment ventilation isolation signals were generated for each unit, Train A of control room ventilation shifted to the make-up mode of operation, and a Train B actuation signal reposi-tioned components of the fuel handling ventilation system. Several balance-of-plant components were also affected. Affected systems were returned to their normal configurations, and appropriate notifications were made. As corrective action, the licensee will evaluate radiation monitors to determine ways to minimize actuations resulting from spurious voltage perturbations. This LER is considered closed.
(Closed) 456/89004-LL: Reactor Trip from Governor Valve Closure due to Defective Turbine Trip Test Switch. On March 6, 1989, Unit 1 tripped from 97% power during a surveillance test of Train B Turbine Trip Relay K640. A faulty test control switch, which is interlocked with the digital electro-hydraulic control system, resulted in the reduction of the turbine governor valve position limits to zero and the subsequent closure of the governor valves.
Decreased steam flow, increased steam pressure, and steam generator shrink effects ensued, and the reactor tripped on low-low levels in the IC and ID steam generators. All protective systems operated as designed, and stable conditions were quickly established. As corrective action, the defective switch will be replaced. Prior to replacement, the defective switch will be bypassed by an electrical jumper during tests involving its use. This LER is considered closed.
In addition to the foregoing, the inspector reviewed the licensee's Deviation Reports (DVRs) generated during the inspection period. This was done in an effort to monitor the conditions related to plant or personnel performance, potential trends, etc.
DVRs were also reviewed to ensure that they were generated appropriately and dispositioned in a manner consistent with the applicable procedures and the quality assurance manual.
No violations or deviations were identified.
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4.
Regional Request Diesel Fuel Oil Reliability In response to an event at Perry Unit 1, a memo concerning reliability of diesel fuel was generated and distributed to all sites. The memo requested information relative to:
1) fuel oil presently used for
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emergency diesel generators, 2) estimated shelf life / stability period, and 3) frequency of fuel oil change-out. The following information was supplied by the licensee:
a.
Type of Fuel Oil:
A blend of 75% #2 and 25% #1 is used.
Receipt of fuel oil is contingent upon ASTM D-975-77 conformance and a maximum sulphur level of.28%.
The blend is utilized to reduce the cloud point temperature to prevent wax formation in the winter months.
b.
Shelf life:
Years. The storage tanks are sampled monthly for particulate contamination, which is an indication of fuel oil breakdown.
Fuel is added to these storage tanks monthly; therefore, fuel oil life in a particular tank is difficult to calculate (see below).
c.
Frequency of fuel oil change-out:
Fuel within the storage tanks is not periodically / routinely changed out. Of course, if the fuel oil in a particular tank shows excessive signs of breakdown, it would be removed.
Storage tanks are/will be cleaned at a 10 year interval. Note that fuel oil is consumed at a rate of about 1400 gallons per diesel generator per month. This equates to about three years to turn over the volume of a 50,000 gallon tank.
The above information was also verbally transmitted to the Region III Technical Support staff.
No violations or deviations were identified.
5.
Follow-up on TMI Action Items I.G.I.3 Training During Low Power Testing-Training and Results This section of NUREG-0737 addresses training on the mechanics of natural circulation flow. Section e.18 Amendment 46 of the Final Safety Analysis Report (FSAR) addresses the licensee's position on operator training for l
natural circulation flow mechanics. The licensee's training program addresses this area during fundamental system training, mitigating core
damage training, and classroom training for the simulator, and simulator training for initial licensed operator training covers natural q
circulation simulation. The inspector's review has identified that the i
licensee has fulfilled its FSAR commitment. This item is considered closed.
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J No violations or deviations were identified.
6.
Unit 2 Boron Dilution Prevention System (BDPS) Inoperability I
On March 12, 1989, at 11:30 a.m., with the unit in Mode 5, the BDPS was declared inoperable after it failed to actuate in response to a
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l spike on a source range instrument channel. The source range instrument spike was caused by welding activity. The licensee then initiated
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. troubleshooting activities to determine why the BDPS had failed to
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operate. The investigation identified that. electrical leads. supplying power to the K6-42 (BDPS) relay had been lifted.
Further investigation
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by the licensee and the resident inspector identified that the leads had been lifted for implementation of temporary alteration 89-2-08. The temporary alteration was installed on February 28, 1989.
Inoperability of the BDPS due to the temporary alteration installation was not discovered until March 12, 1989.
On March 13, 1989, at 1:30 p.m., the temporary alteration was removed to restore the BDPS to operation. The root cause of this event was an inadequate tecnnical review by the technical staff ' engineer preparing the temporary alteration and. an inadequate onsite review (OSR) to evaluate the technical scope of the proposed temporary alteration. The review failed to identify that inoperability of the BDPS would result if the proposed temporary alteration were implemented. The 10 CFR 50.59 evaluation erroneously stated that the affected equipment was not required for Mode 5.
The BDPS system is required for Mode 5 as a function of the source range
~ instrumentation. -The licensee takes credit in the FSAR for the BDPS operation.
In the event of a boron dilution transient, the nuclear
. instrumentation source range in conjunction with the flux doubling detection circuit will detect a doubling of the neutron flux. This information is sent to the solid state protection system which automatically initiates isolation valve movement to terminate the transient. An alarm is sounded at the time for plant operators to indicate that flux doubling has occurred and isolation valve movement started. Credit is taken for the instrumentation to provide for operator alert and for automatically initiating isolation valve movement.
Discussion with technical staff personnel identified that the engineer preparing the temporary alteration should have been aware of the solid state' protection system (SSPS) panel wiring scheme for the BDPS and auxiliary feedwater system relays.
Failure to perform a complete review of appropriate wiring lists and diagrams is also contrary to the requirements of procedure BwAP 320-1, " Temporary Alterations."
The technical review of the temporary alteration is considered by the NRC to be an integral part of the 10 CFR 50.59 review process. Therefore, the 10 CFR 50.59 review is also considered to be inadequate. To perform an adequate 10 CFR 50.59 review, the licensee has to assess the technical adequacy of information submitted for review to ensure that the proper safety perspective is reviewed.
Failure to perform a proper technical review and subsequent 10 CFR 50.59 evaluation is considered a violation of 10 CFR 50.59 (457/89009-01(DRP)).
One violation was identified.
7.
Inadvertent Safety Injection (SI)
On April 16,1989, at 4:40 p.m., during a plant heat-up and pressurization from Mode 5 to Mode 3, following a one-week maintenance outage for turbine governor valve and inverter 113 repair, RCS pressure increased above the P-11 setpoint of 1930 psig.
However, the steamline pressure was less than 640 psig.
Hence, an inadvertent SI was initiated
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'because P-11 unblocks the steamline pressure SI logic. The event also caused a temperature transient, a 105"F increase in 36 minutes, in the pressurizer; this heat-up rate exceeded the pressurizer Technical
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Specification heat-up rate of 100 F in a one-hour period. An engineering review was performed to evaluate the effect on system operation. All Emergency Core Cooling System (ECCS) equipment operated as designed, and 5000 gallons of water was injected into the RCS. Unit heat-up and pressurization is a normal evolution for which all op2 rating personnel receive formal training, including simulator time. On the previous shift, the nuclear station operator (NS0) had taken actions to correct an RCS temperature / pressure condition in which the pressure was below the target value for the existing temperature. At the time of the shift turnover (3:00 p.m.), due to actions taken by the previous shift NS0 to increase pressure toward the target pressure, pressure was increasing and temperature and pressure were close to the target values. By 4:40 p.m.
the pressure had reached 1935 psig with RCS temperature at 500 F and a corresponding steam generator pressure of about 639 psig. The target RCS temperature for 1930 psig is 525 F.
Other activities also in progress during the plant heat-up and pressurization associated with the feedwater station and a data recorder may have diverted the NS0's attention from the increasing RCS pressure.
The caution in procedure BwGP 100-1, " Plant Heatup," states, "To prevent an inadvertent SI, steamline pressure must be greater than 640 psig, prior to exceeding the P-11 setpoint (1930 psig RCS pressure), as the low steam line pressure (640 psig) and low pressurizer pressure (1829 psig)
SI signals automatically unblock, Pressures must be maintained above their respective setpoints." The procedure, for which the NSO received formal training, was located on the NS0's desk prior to the event.
Discussions between the resident inspector and licensee personnel identified the cause of this event to be personnel error. The event could have been prevented if the operations personnel had taken action to control system pressure. The pressure increase rate was of a long enough duration that the operating personnel had sufficient time to take the appropriate actions to prevent the SI actuation.
Procedure BwAP 300-1, " Conduct of Operations," states, "All operating personnel must be alert and remain within their immediate areas of responsibility until properly relieved and be responsible for monitoring the instru-mentation and controls located within their areas.
They are responsible for taking timely and proper actions to ensure safe operation of the facility." It is the responsibility (NSO, shift control room engineer [y of all operating personnel on shift SCRE], and shift engineer) to ensure that plant instrumentation and controls are adequately monitored to facilitate safe operation of the facility. This failure to adequately monitor plant instrumentation and take proper corrective actions to control RCS pressure, in order to prevent the inadvertent SI, is considered a violation of 10 CFR 50, Appendix B, Criterion V (456/89009-01(DRP)).
One violation was identified.
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8.
Contaminated Equipment During the inspection period, licensee radiation protection personnel informed the resident inspectors that they had discovered unexpected contamination on the reactor coolant pump balancing gear. The equipment, which had been shipped from a Westinghouse facility in Pennsylvania for use at Braidwood, was not labeled as radioactive. The licensee discovered the contamination during surveys prior to returning the l
equipment to Westinghouse. Two bags of contaminated gear (mostly i
electrical cabling) had been collected, one containing fixed contamination of 30,000 dpm/100 cm2 and the other containing smearable contamination of 1,000 dpm/100 cm. None of the equipment in the two r
bags had been used in radiologically controlled areas at-Braidwood, so the equipment must have been contaminated before it was received at Braidwood.
Upon discovery, all contaminated gear was handled and stored under appropriate radiological controls. All other equipment from the same shipment was surveyed and found to be uncontaminated. There was no evidence of personnel contamination from the equipment. Westinghouse later determined that the equipment had been used at the H. B. Robinson plant prior to being shipped to Braidwood; the equipment was apparently contaminated when returned to Westinghouse by Robinson.
Region III radiation protection personnel have informed Region I and Region II for follow-up.
No violations or deviations were identified.
9.
Operational Safety Verification (71707)
During the inspection period, the inspectors verified that the facility was being operated in conformance with the licenses and regulatory requirements and that the licensee's management control system was effectively carrying out its responsibilities for safe operation.
This was done on a sampling basis through routine direct observation
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of activities and equipment, tours of the facility, interviews and
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discussions with licensee personnel, independent verification of safety system status and limiting conditions for operation action requirements (LC0ARs), corrective action, and review of facility records.
On a sampling basis the inspectors daily verified proper control room l
staffing and access, operator behavior, and coordination of plant activities with ongoing control room operations; verified operator adherence with the latest revisions of procedures for ongoing activities;
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verified operation as required by Technical Specifications (TS);
i including compliance with LC0ARs, with emphasis on engineered safety
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features (ESF) and ESF electrical alignment and valve positions; i
monitored instrumentation recorder traces and duplicate channels for abnormalities; verified status of various lit annunciators for operator understanding, off-normal condition, and corrective actions being taken; examined nuclear instrumentation (NI) and other protection channels for proper operability; reviewed radiation monitors and stack monitors for
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abnormal conditions; verified that onsite and offsite _ power was available as required; observed the frequency of plant / control room visits by the-station managar,. superintendents, assistant operations. superintendent,
- and other managers; and observed the Safety Parameter Display System (SPDS) for operability.
During tours of accessible areas of the plant, the inspectors made note of general plant / equipment conditions, including control of activities in progress (maintenance / surveillance), observation of shift turnovers, general safety: items, etc.
The specific areas observed were:
Engineered Safety Features (ESF) Systems L
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Accessible portions of ESF systems and components were inspected to verify:
valve position for proper flow path; proper alignment of-power supply breakers or fuses (if visible) for proper actuation on
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an initiating signal; proper removal of power from components if-required by TS or FSAR; and the operability of support' systems essential to system actuation or performance through observation of
- instrumentation and/or proper valve alignment. The inspectors also
. visually inspected components for leakage, proper. lubrication, cooling water supply, etc.
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Radiation protection Controls-The inspectors verified that workers were following health physics r
procedures for dosimetry, protective clothing, frisking, posting, etc., and randomly examined radiation protection instrumentation for use, operability, and calibration.
- Security The inspectors, by sampling, ' verified that persons in the protected area (PA) displayed proper badges and had escorts if required; vital areas were kept locked and alarmed, or guards posted if' required; and personnel and packages entering the PA received proper search and/or monitoring.
- Housekeeping and Plant Cleanliness The-inspectors monitored the status of housekeeping and plant cleanliness for fire protection, protection of safety-related equipment from intrusion of foreign matter and general protection.
The inspectors also monitored various records, such as tagouts, jumpers,
'shiftly logs and surveillance, daily orders, maintenance items, various chemistry and radiological sampling and analysis, third party review results, overtime records, QA and/or QC audit results and postings
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required per 10 CFR 19.11,
During the inspection the licensee provided information on several issues
- raised by the inspectors. The following is a summary of those topics considered:
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Management t6urs of the plant.
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- The licensee has developed and implemented a form to document yb observations re' quiring attention'that are identif Sd during tours.
of.the plant by management personnel.
L Control room panel cleaning policy.
. The licensee has implemented a plan for assigned cleaning stations with responsibilities assigned to specific personnel who work in the control room.
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Residual Heat Removal (RHR) pump operation without
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(CC) to the shaft seals.
The inspectors raised this issue wnen 1.; was found that-on an
. occasion, the RHR pumps had been operated during cold shutdown,
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above the shutoff head without CC.
Since the pump seal is a fission product boundary and the pump could be called upon for long periods-of operation under either normal or accident conditions, the inspector. questioned the prudency of this action. The licensee provided corporate and manufacturer's documentation showing that the RHR pumps could be operated for up to 2.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> while in the SI mode and above the shutoff. head.
Each 4 the responses appear to be acceptable and are considered closed.
-The < effectiveness will be monitored in the future as part of the routine inspections.
No violations or deviations were identified.
10. Monthly Maintenance Observation (62703)
Station maintenance ~activitic, affecting the safety-related systems and components listed below were observed / reviewed to ascertain that they were conducted in accordance with approved procedures, regulatory guides and industry codes or standards, and in conformance with Technical Specifications.
The following 1tems were considered during tisis review:
the limiting cor.ditions for operation were met while components or systems were removed from and restored to service; approvals were obtained prior to initiating the work; activities wc 4 accomplished using approved procedures and were inspected as.mplicable; functional testing and/or calibrations were performed prior co re*;st'ng components or systems to service; quality control records we 1 al ntained; activities were accomplished by qualified personnel; pt :
and materials used were properly certified; radiological controls were implemented; and fire prevention controls were implemented. Work requests were reviewed to determine the status of outstanding jobs and to assure that priority is assigned to safety-related equipment maintenance which may affect system performance.
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The following maintenance activities were observed and reviewed:
Unit 0 Changeout of' spent fuel storage racks to.high. density-racks and drag L-testing of racks.
Unit 1 IFWOO9 valves hydraulic repair / refurbishment.
Instrument Bus 111 inverter repair.
Instrument Bus 113 inverter repair.
' Unit 2 Replace ent of seal injection filter 2A.
'2TI-411B Delta-T, T AVG Loop 2A troubleshooting / repair.
The inspectors monitored the licensee's work in progress and verified that'it was being performed in accordance with proper procedures, and
- approved work packages, tnat 10 CFR 50.59 and other applicable drawing updates were made.and/or planned, and,that operator training was l'_.
conducted in a reasonable period of time.
No violations or deviations were identified.
11. Monthly Surveillance Observation (61726)
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.The inspectors observed surveillance testing required by Technical Specifications during the inspection period and verified that testing was performed in accordance with adequate procedures, that' test
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instrumentation was calibrated, that limiting ~ conditions-.for operation were met', that removal and restoration of the affected components were accomplished, that results conformed with. Technical Specifications and procedure requirements and were reviewed by personnel other than the
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individual directing the test, and that any deficiencies identified during the testing were properly reviewed and ' resolved by appropriate management persorrel.
l The inspectors also witnessed portions of the following test activities:
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Unit 1 BwVS 3.1.1-5, Incore-Excore Axial Flux Quarterly Calibration.
BwIS 3.1.1-223, Quarterly Excore/AFD Calibration.
Bw05 8.1.1.2.a-1, IA Diesel Generator Operability Monthly Surveillance.
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-. : s BwIS 3.1;1-336, Steam Generator ID Level Analog Operational Test and Channel Verification.
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BwlS'3.3.6-301, Refueling Water Storage Tank Level Analog Operational Test and Channel. Verification.
' Unit 2 BwVS 2.3.5-1, Reactor Coolant-System Flow Measurement.
BwVS 0.5-3-DO.1, ASME Requirement for. Test of the Diesel Oil Transfer System.
Bw0S 8.1.1.2.a-2, 28 Diesel Generator Monthly Operability.
No violations or deviations were ' identified.
12.
Resumption of-Normal Operations After the Strike by Security Guards (92712)
This issue was addressed'in inspection reports 456/89002(DRP);
457/89002(DRP) and 456/89005(DRP); 457/89005(DRP). On March 18, 1989,
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the striking. security guards of Local 228 of the United Plant Guard Workers of America (UPGWA) and the management of Burns Security reachee a. settlement. On March 23, 1989, the striking guards returned to work.
The ~ resident inspectors verified that the transition from the strike condition to normal. operations went without incident.
'In addition, a Region III security specialist was dispatched to the site to monitor the transition. The results are documented in inspection report 456/89010(DRSS); 457/89010(DRSS).
No violations or deviations were identified.
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'13.
Systematic Appraisal of Licensee Performance (SALP) Meeting On April 26, 1989, a public meeting was held at the Braidwood site between members of the NRC staff, headed'by the Regional Administrator, and the licensee, headed by the President of Commonwealth Edison Company, to discuss the SALP 8 Report. One representative of the public was present. The licensee acknowledged the report's ratings and stated that a written response was forthcoming. The details of the assessment are contained in the SALP 8 Board Report (456/89001; 457/89001) and in related correspondence.
No -violations or deviations were identified.
14. Augmented Inspection Team (AIT)
From April 25 to April 28, 1989, a five-member AIT, headed by the Acting Chief of DRP Branch 1, was at the Braidwood site to investigate a concern involving worker attentiveness.
The concern arose a week earlier when an NRC regional inspector observed two individuals who appeared inattentive.
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.The AIT thoroughly investigated the two incidents, interviewed a
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cross-section of the station work force to ascertain the overall station
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- attitude toward inattentiveness, and reviewed documents containing.the L
licensee's policy and directives associated with worker attentiveness.
The results and recommendations of the AIT inspection are contained in Inspection heport 456/89014; 457/89014.
No violations or deviations were identified.
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g 15.
Training Effectiveness (41400, 41701)
The effectiveness of training programs for licensed and non-licensed personnel'was reviewed by the inspectors during the witnessing-of the
' licensee's performance of routine surveillance,~ maintenance, and
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operational activities and during the review of the licensee's response-to events which occurred during the-inspection period.
Personnel appeared to be knowledgeable of the tasks being performed, and nothing.
was. observed which indicated any ineffectiveness of. training.
No violations or deviations were identified.
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16.
Report Review During the inspection period, the inspector reviewed the licensee's Monthly Performance Report for March 1989. The inspector confirmed that i-the information provided met the requirements of Technical Specification 6.9.1.8 and. Regulatory Guide 1.16.
The inspector also reviewed the licensee's Monthly Plant Status Report for February and March 1989,.and. the meeting' notes of the Braidwood Corporate Overview meetings held'on March 9 and' March 23,~1989.
No violations or deviations were identified.
17.
ExitInterview(30703).
The. inspectors met with the licensee representatives denoted in paragraph:1 during the inspection period and at the conclusion of the inspection on April 27, 1989..The inspectors summarized the scope and results of the inspection and discussed the likely content of this inspection report. The licensee acknowledged the information and did not indicate that any of the information disclosed during the inspection could be considered proprietary in nature.
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