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Insp Repts 50-277/93-15 & 50-278/93-15 on 930615-0802. Violations Noted.Major Areas Inspected:Plant Operations, Maint & Surveillance,Engineering & Technical Support,Plant Support & Physical Security
	ML20024J030
Person / Time
Site:	Peach Bottom
Issue date:	08/19/1993
From:	Anderson C; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:	;
Shared Package
ML20024J025	List: IR 05000277/1993015;
References
	50-277-93-15, 50-278-93-15, NUDOCS 9308310170
	Download: ML20024J030 (80)
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U. S. NUCLEAR REGULATORY COMMISSION

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REGION I

DOCKET / REPORT NOs: 50-277/93-15 50-278/93-15 LICENSE NOs:

DPR-44 i

DPR-56

LICENSEE:

Philadelphia Electric Company

Peach Bottom Atomic Power Station P. O. Box 195 Wayne, PA - 19087-0195 d

FACILITY NAME:

Peach Bottom. Atomic Power Station Units 2 and 3 DATES:

June 15 - August 2,1993 INSPECTORS:

B. S. Norris, Acting Senior Resident Inspector F. P. Bonnett, Resident Inspector R. K. Imrson, Resident Inspector

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APPROVED BY:

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C. J. Andelson, 8hief Date Reactor Projects Section 2B i

j Division of Reactor Projects

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. 9308310170 930823

' DR ADOCK 05000277 i

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IVE SUMMARY Atomic Power Station on Report 93-15 js observed. The Unit 3 refueling outage was well bnnel responded appropriately to a loss of vacuum

f the off-gas recombiner system (Section 2.0). One lce was noted during recovery from a planned power mispositioned (Section 1.1). The mispositioning of of Technical SpeciGcation 6.8.1 with respect to janresolved item was identined regarding the program e Senior Reactor Operators active (Section 6.4, URI lE bserved during a hydraulic control accumulator level j

jeficiencies are considered a violation of Technical d-02). An unresolved item was identi5ed that seismic

'eral of the safety-related 480vac breakers. The priate. This issue will remain unresolved pending w by the NRC (Section 6.3, URI 93-15-03).

MPPORT

in response to issues regarding non-condensable gas

'rumentation. The short-term compensatory measures nodi 5 cation designed to ensure reactor vessel water

)nsistent with the Bulletin 93-03 (Section 4.1). Also,

' ped to backGil the reference leg of the reactor vessel The inspectors noted that this procedure was well

pe decision to install the hardware modification in the lementation of this backfill procedure was considered nerally good during the period, however, multiple rker procedures were observed during the outage, and removal of the Protective Clothing, improper to log one person out of a controlled area (Section Slation of Technical Specification 6.8.1 with respect ii

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TABLE OF CONTENTS Page EXECUTIVE SUMMARY ii

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1.0 PLANT OPERATIONS REVIEW

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Unit 2 Mispositioned Control Rod......................... 1 l

1.1 l.2 Site Hurricane Readiness............................... 2

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2.0 FOLLOW-UP OF PLANT EVENTS............................ 3 3.0 U NIT 3 O UTA G E....................................... 4 3.1 Backg ro und....................................... 4 3.2 Replacement of Leaking Fuel and Fuel Inspection

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3.3 U ni t 3 S tart-up..................................... 6 3.4 Conclusion

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4.0 ENGINEERING AND TECHNICAL SUPPORT

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4.1 Response to NRC But sin 93-03.......................... 7 4.2 Backfill of Reactor Vessel Water Level Instrument Reference Irg

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l 5.0 SURVEILLANCE TESTING OBSERVATIONS..................... 8 6.0 M AINTENANCE ACTIVITY OBSERVATIONS.................... 9

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6.1 E-3 3 Bu s O u tage...............................

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.. 9 6.2 D rywell Tou r.....................................

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l 6.3 Operability of 480 VAC Circuit Breakers....................

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6.4 Limited Senior Reactor Operators

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i 7.0 R A DIOLOGICAL CONTROLS..............................

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8.0 PHYSICAL SECURITY

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9.0 PREVIOUS INSPECTION ITEM UPDATE

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10.0 M ANAG EM ENT MEETINGS...............................

i 10.1 Performance Review

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10.2 Ex i t M eeti n g.....................................

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DETAILS

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1.0 PLANT OPERATIONS REVIEW (71707)*

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The inspectors monitored plant activities by directly observing safety significant activities and equipment, touring the facility, and interviewing and discussing items with licensee personnel.

The inspectors independently verified safety system status and Technical Specification (TS)

Limiting Conditions for Operation (LCO), reviewed corrective actions, and examined facility l

l records and logs. The inspectors performed 19 hours2.199074e-4 days 0.00528 hours 3.141534e-5 weeks 7.2295e-6 months of deep backshift inspection of the facility.

Unit 2 operated at essentially 100% power for the entire inspection period. The unit did not experience any major transients or enginected safety feature (ESP) actuations. During the report period, the inspectors evaluated licensee staff and management response to plant equipment problems to verify that the licensee had identified the root cause, implemented appropriate corrective actions, and made the required notifications. The inspectors observed control room operations and noted that supervision maintained good oversight of activities and responded appropriately to equipment problems.

One exception to the otherwise good operator

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l performance was noted during the recovery from a planned power decrease when a control rod was mispositioned (Section 1.1).

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At the beginning of the inspection period, Unit 3 was operating at about 75% power. The licensee shutdown the unit on July 4, for a limited refueling outage to replace three leaking fuel bundles (Section 3.0). The licensee restarted the Unit on July 15, and achieved 100% power on July 19 (Section 3.4). At the conclusion of the period, Unit 3 was manually scrammed.

Blown fuses in the off-gas recombiner system led to loss of the system, which caused a loss of condenser vacuum. After determining the cause of the low condenser vacuum, the licensee

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restarted Unit 3 within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of the shutdown. At the end of the period, the unit had returned to approximately 100% power.

1.1 Unit 2 Mispositioned Control Rod (71707, 92701)

On June 24,1993, at 3:23 a.m., while Unit 2 recovered from a planned load drop, the licensee discovered that control rod 14-31 had been mispositioned for approximately 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . The reactor was operating at 60% power when the mispositioning occurred. The control room staff i

immediately entered off-normal procedure ON-122, "Mispositioned Control Rod," and i

repositioned the control rod to its correct position. Shift management notified the Plant and Operations Managers, and initiated a Reportability Evaluation / Event Investigation Form (RE/EIF). Due to the unique nature of the event and its potential effect on reactor safety, the inspector utilized the NRC's Human Performance Investigation Process (HPIP) for follow-up inspection.

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Prior to the mispositioning, the licensee had reduced reactor power to 60% power to facilitate repair of a high pressure coolant injection (HPCI) system check valve. During performance of

The inspection procedure frorn NRC Ma - ! Chapter 2515 that the inspectors used as guidance is parenthetically

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the repair, the maintenance job leader requested that reactor power be reduced to further reduce l

job site radiation exposure. The Shift Supervisor (SSV) directed the reactor operator (RO) to l

insert the next control rod listed on Table 1 of GP-3-2 Appendix 1, " Unit 2 Shutdown Rod Insertion Sequence Instructions." The RO selected control rod 14-31, which was at notch

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position 34, and fully inserted it.

When maintnenance was completed, the SSV directed the RO to withdraw rod 14-31 to restore reactor power. The SSV did not refer to General Procedure (GP)-5, " Power Operations," prior to the power increase, as required. GP-5 directs power increases to be performed in accordance with Reactor Engineering procedure RE-31, " Reactor Engineer Startup/Imad Drop Instruction."

Exhibit 9 of RE-31, " Control Rod Position Data Sheet," had the last position of control rod 14-31 listed as position 34. Contrary to RE-31, Exhibit 9, the RO withdrew control rod 14-31 l

to its fully withdrawn position. The control rod position was not documented in RE-31, as i

required.

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The mispositioning was identified about two hours later by the Reactor Engineer (RE) when he requested a computer generated core evaluation report. The report was automatically aborted due to the mispositioning of control rod 14-31 which cause an asymmetrical rod pattern. The

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RE performed a rod position verification and identified that control rod 14-31 was at position l

48 vice position 34.

l The inspector discussed these issues with the operation's manager who agreed with the findings and implemented appropriate interim corrective actions. These actions included meetings to j

heighten personnel awareness and a procedure revision to GP-5. The inspector informed the i

licensee that the above constitutes a violation of TS 6.8.1 (Violation 50/277 & 50-278/93-15-01).

l 1.2 Site Hurricane Readiness

Following the aftermath of Hurricane Andrew, the inspector ieviewed the licensee's procedures for hurricanes and severe weather. Specifically, the review focused on preparedness for a severe l

storm, reliability of off-site communications during a storm, and compensatory measures for l

buildings and outside equipment not designed to withstand a storm. The inspector noted that the l

licensee does not have an integrated severe weather response procedure. However, the licensee has taken adequate measures in the past to prepare the site for approaching storms. The l

inspectors reviewed the licensee's readiness for a severe winter storm last March (NRC l

Inspection Report 50-277 & 278/93-03). Specific procedures currently exist to mitigate storm related effects (i.e., flooding). The licensee's Emergency Plan states general guidance regarding

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event classifications for hurricanes. The inspector discussed the issue with licensee management l

who indicated that development of a severe weather procedure was under review. The licensee has been cleaning up storage areas and removing loose articles to enhance their hurricane preparedness. The inspector had no further questions.

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2.0 FOLLOW-UP OF PLANT EVENTS (71707, 93702)

During the report period, one event occurred. On July 30,1993, at 3:27 a.m., the licensee manually scrammed the Unit 3 reactor after a loss of condenser vacuum. The licensee was performing Special Procedure (SP)-1438, " Procedure to Purge and Vent Process Lines, Condensing Pots, and Flow Transmitters for Steam Flow to the Recombiner," to vent the transmitter for steam flow to the off-gas recombiner. This procedure had been performed previously due to drifting of the transmitter. The recombiner steam inlet valve shut following the removal of a jumper used to prevent a low steam flow isolation. The operating crew noted the change in valve position and the decreasing vacuum. The operators entered the Operating Transient (OT) procedure, OT-106, "less of Condenser Vacuum," and attempted to recover the off-gas system. The Reactor Operator initiated a manual scram when the first low condenser vacuum alarm actuated. Following the scram, reactor water level decreased, as expected, to -20 l

inches and caused a primary containment isolation system (PCIS) group II/III isolation. The operators restored level to its normal band using the feed pumps, initially, and maintained i

reactor level and pressure control using the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) systems. The operators used the appropriate Transient Reactor

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Implementing Procedures to stabilize the plant. All engineered safety features responded as l

expected. The NRC was notified via the ENS.

The licensee performed an investigation of the cause for the off-gas recombiner system isolation.

They found that two blown fuses in the low flow isolation circuits had caused the isolation. The two circuits have to actuate for an isolation to occur. The licensee's Technical Group concluded that one fuse had been blown prior to the event and the other fuse blew when the jumper was I

removed. To confirm this conclusion, the licensee's I&C technicians replaced the fuses and satisfactorily tested the recombiner circuitry.

The inspector monitored the licensee's troubleshooting activities and attended the post-scram critique. The licensce fully reconstructed the event, identifying strengths and weaknesses. A noted strength was the Shift Manager's pre-evolution brief conducted prior to performance of the SP-1438 evolution. This brief enhanced the shift's ability to respond to the event. The inspectors evaluated the response to the event to verify that the licensee had identified the root cause, implemented appropriate corrective actions, and made the required notifications.

The inspector discussed the event and corrective actions with licensee management. The licensee

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has developed a modification to change the piping configuration of the flow transmitter to minimize the need to vent the transmitter. The modification is planned for the upcoming i

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refueling outage. As a short-term corrective action, the licensee will revise the SP procedure

to perform "before and after" fuse continuity checks to ensure fuse continuity. Based on the above observations and discussions, the inspector concluded that.the licensees actions were acceptable.

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The licensee commenced start-up on July 30, at 2:47 p.m. The inspectors observed the start-up due to the high Xenon concentration that still existed in the reactor after the scram. The operations staff performance was deliberate, cautious, and observed safe operating practices.

3.0 UNIT 3 OUTAGE An unplanned Unit 3 mid-cycle outage began on July 6,1993, to replace two known leaking fuel bundles. The reactor was shutdown on July 4 to allow sufficient time for decay of radiological components. Major activities for the outage included: (1) fuel replacement and sipping of the fuel bundles within the two affected cells, (2) repair of the 3A traversing in-core probe (TIP)

l indexer, (3) replacement of three control rod position indication probes (PIPS), (4) E-33 bus l

outage work, and (5) repair of a packing leak on an inboard main steam isolation valve (MSIV).

l The ingters attended various meetings during the outage to evaluate the effectiveness of the licensee's planning, coordination, and communication efforts. The inspectors observed work activities on the refueling floor, in the drywell, and at E-33 bus outage locations to assess the licensee's work practices and effectiveness. The inspectors also reviewed the Limited Senior Reactor Operator's (LSRO) program for maintaining their licenses active.

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3.1 Background The licensee detected a fuel leak on Unit 3, in control cell 46-31 in May 1992. At that time, the licensee fully inserted control rod 46-31 and the three symmetrical control rods for that rod group to suppress the fuel leak. In January 1993, the licensee performed flux tilt testing and confirmed a suspected fuel leak in the center control cell (30-31). The licensee inserted the control rod to its full-in position to suppress the leak.

In early May 1993, the licensee performed a load drop for maintenance and routine rod pattern adjustment. One of the repositioned control rods was in the vicinity of the two known fuel leaks. Following the rod pattern adjustment, the licensee noted a step increase in off-gas activity; subsequently, the rod group was reinserted to its original position. Based on analysis, the licensee determined that one or both of the leaking fuel bundles had degraded. The analysis measured reactor coolant activity for gaseous fission products, specifically three krypton isotopes and three xenon isotopes. The licensee performed a second load drop on May 22,1993, to perform flux tilt testing for leaking fuel bundles.

Based on the test results and vendor recommendations, the licensee fully inserted a control rod to further reduce the power generation in the leaking fuel bundle and minimize additional degradation of the leaking fuel pin. This established an asymmetrical rod pattern, which reduced the maximum power limit of the reactor to 95 %.

At the beginning of June 1993, a second adjacent control rod was inserted to shadow the leaking fuel bundle, resulting in an early end-of-cycle coastdown. Even though there were seven control rods inserted, off-gas levels continued to increase at a steady rate. The licensee compared the increasing off-gas activity rate to a similar problem which occurred at Limerick Generating Station, Unit 2. The comparison indicated that Unit 3 off-gas activity was increasing at the same

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rate as Limerick. On June 28,1993, the licensee decided to perform a mini-outage to replace

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the leaking fuel bundles.

3.2 Replacement of Leaking Fuel and Fuel Inspection The Unit 3 reactor uses barrier fuel in a control cell core design. This design uses a fixed group of control rods to control reactivity during power operations. All other control rods are normally fully withdrawn. A control cell has four low-reactivity fuel bundles so that control rod I

motion occurs adjacent to low-power fuel which offsets the effects of reactivity changes during power operations. The barrier fuel design consists of a layer of pure zirconium on the inside of the clad.

The layer reduces stresses and chemical attacks caused by pellet-to-clad j

interactions.

The licensee's Fuel Management Group (FMG) selected two barrier fuel bundles from the l

periphery of the core to replace the leaking fuel in the barrier cells. They also identified four

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replacement fuel bundles for the periphery from old fuel stored in the spent fuel pool. The old bundles were non-barrier type fuel that had been removed from the reactor during the previous fuel cycle. The old fuel was sipped and visually inspected for integrity. The FMG closely matched the replacement fuel's reactivity requirements and performed a safety analysis as required by 10 CFR 50.59.

l Core alterations to replace the leaking fuel bundles began on July 9,1993. The licensee

" sipped" or sampled all eight fuel bundles in control cells 30-31 and 46-31 for leaking fission gases and determined that two bundles had extensive fuel leaks. The two bundles were replaced.

A third bundle was found to have a small pin hole leak. The licensee revised the Core Component Transfer Authorization Sheet and safety analysis report and replaced the third bundle. All other fuel bundles were acceptable. A partial core verification in a three-by-three array around each cell was performed. Core alterations were completed about 2:00 p.m., on July 10,1993.

The inspector reviewed the safety analysis report and discussed it and the refueling evolution with the lead Reactor Engineer. The RE fully explained the engineering process for selecting the replacement fuel bundles, sipping the affected fuel bundles, and determining the leaking bundles. The inspector had no further questions.

The licensee disassembled and inspected the leaking fuel bundles. The inspection identified bulging and numerous splits in the clad of a several fuel pins. A split about 40 inches long displayed extensive washout of the pin's fuel pellets. In the three bundles, the licensee found no evidence of debris induced fretting. The licensee was not able to determine the primary failure mechanism.

The inspector observed portions of the fuel inspection and examined the cracked fuel pin.

Besides the cracks already noted, there was evidence of cracks that could have continued to propagate had the fuel remained in operation. The inspector discussed the possible cause of the

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damage with a vendor representative. The fuel pin displ'ayed evidence of extensive secondary hydriding; i.e., when debris in the coolant becomes trapped in the fuel bundle and frets a small

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defect into the clad of the fuel pin. Water enters the fuel pin and flashes to steam. Hydrogen

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from the steam attacks the pure zirconium barrier inside the clad, embrittling the barrier and causing the clad to bulge until a crack begins. Heat causes the crack to propagate, unzippering the fuel pin and allowing the flow of coolant past the pin to washout the fuel pellets.

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The inspector maintained open communications with the licensee's reactor engineering and fuel

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management groups from the time the fuel leak was detected until the fuel was inspected. The inspector concluded that the licensee's approach and strategy toward detecting and suppressing further degradation of the fuel were sound and well-planned. The inspector agreed with the licensee that continued plant operations until the fall outage would have further degraded the fuel resulting in additional radiological problems. The inspector had no further questions.

3.3 Unit 3 Start-up l

The licensee completed reactor reassembly on July 14, 1993, and placed the mode switch in

" start-up" at 1:30 p.m. on July 15. The licensee achieved criticality at 5:21 p.m.

Power ascension continued through the weekend, and the reactor reached 100% power early on July 19. The inspector observed the control room activities during the preparation for start-up, approach to criticality, and reactor heat-up. Prior to ccmmencing rod pull, the shift supervisor held a crew briefing covering expected plant response and contingency plans. The start-up was orderly and well-controlled. The RO exercised caution and safe operating practices throughout the start-up.

3.4 Conclusion Early detection and suppression of the leaking fuel avoided more severe degradation of the fuel.

Off-gas rad levels were maintained reasonably low due to the consenative action taken by the licensee. Staff and management were well informed, emerging problems were promptly raised and addressed, and coordination among working groups was observed to be excellent. Overall, the outage was well planned, handled professionally, and well managed.

4.0 ENGINEERING AND TECHNICAL SUPPORT (37700, 71707)

The inspectors routinely monitor and assess licensee support staff activities. During this inspection period, the inspectors focused on the licensee's response to issues involving noncondensable gas buildup in the reactor vessel water leve.. -

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4.1 Response to NRC Bulletin 93-03 In May 1993, the NRC issued Bulletin (BL) 93-03, " Resolution of Issues Related to Reactor Vessel Water level Instrumentation in BWRs." This bulletin provided information to licensees regarding reactor vessel water level indication errors that could occur during plant depressurization, and requested licensees to implement specific short-term compensatory measures during cooldown operations. Additionally, BL 93-03 requested licensees to develop hardware modifications to ensure long-term reliability of the reactor vessel water level instrumentation. The inspector reviewed the licensee's implementation of the short-term compensatory actions and planned long-term actions relative to BL 93-03.

The short term actions requested by BL 93-03 included: enhanced monitoring of the reactor l

vessel level instruments, control of evolutions that could potentially drain the reactor pressure vessel during cooldown, and augmented operator training regarding this type of event. The licensee revised the Normal Shutdown Procedure (GP-3) to require enhanced monitoring of the l

reactor vessel water level instrumentation during plant depressurization and review of activities

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that could potentially drain the reactor pressure vessel. The licensee issued a required reading package and a simulator exercise guide (SEG) to train the operators to recognize and properly respond to this type of event. The inspector reviewed the SEG and determined that the exercise scenarios adequately trained the operators to recognize and respond to this type of event. The inspector also observed an operating crew respond to this event in the simulator and noted that the operators' response to the scenario was appropriate. The inspector had no further questions in this area.

The licensee is developing a hardware modification to ensure long-term reliability of the reactor vessel water level instrumentation. This modification is designed to prevent the buildup of noncondensable gasses in the reactor vessel water level instruments' reference legs by providing a constant supply of makeup water. The licensee has commenced installation of components for this modification and intends to complete installation within the time period specified in BL 93-03. The inspector is continuing to follow the licensee's actions on this matter.

4.2 Backfill of Reactor Vessel Water Level Instrument Reference Leg NRC Inspection Report 92-07 discussed problems that occurred on the Unit 2,2B condensing chamber which led to a plant shutdown. The report also noted that the licensee installed a temporary modification to monitor condensing chamber temperatures to gather data in support of further engineering analysis. The licensee also developed acceptance criteria for maximum deviation between level instrument channels. During this inspection period, the inspector reviewed the licensee's evaluation of this data and response to an indicated level divergence on the Unit 2B instrument.

The inspector interviewed the responsible System Manager and noted that the licensee was closely monitoring Unit 2 condensing chamber performance (temperature and level). They used temperature data to predict the condensing chamber performance and potential for level

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divergence on the 2A and 2B level instruments.

The inspector noted that the hcensee

appropriately used the condensing chamber temperature data to alert operating personnel of the potential for indicated level divergence. The inspector concluded that the licensee demonstrated j

strong technical support of plant operations in this area.

The licensee developed SP-1444, "Backfilling of "2B" Condensing Chamber During Power Operations," to provide a method of restoring reference leg inventory while at power. The procedure was critically reviewed by the Plant Operations Review Committee (PORC), and the Independent Safety Engineering Group (ISEG). The inspector reviewed the procedure and determined that it was technically sound, contained appropriate cautions and warnings, and established appropriate controls for the performance of and restoration from the evolution. The inspector also noted that the appropriate Technical Specifications were referenced in the procedure. The inspector had no further questions regarding the procedure.

The inspector observed the licensee's preparations for conducting this procedure. The licensee performed several thorough pre-evolutionary briefs.

Additionally, the licensee identified potential events that could occur during this procedure and conducted simulator training on these i

events prior to performing the evolution. The inspector observed the performance of this evolution and noted that procedural adherence, command and control, and communications were excellent.

The inspector concluded that the licensee's planning, training, briefing, and j

performance of this evolution was a significant strength.

The inspector discussed with licensee management the potential risks involved with repeated use j

of this procedure over an extended time period. During the procedure, a wide range level instrument (and associated safety functions) and an excess flow check valve are inoperable. The licensee indicated that this issue was under review and would impact their schedule for installing the permanent hardware modification. After the end of the reporting period, the licensee

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decided to shutdown and install the permanent modification. The inspector was satisfied with i

this response and will continue to review licensee actions in this area.

5.0 SURVEILLANCE TESTING OBSERVATIONS (61726, 71707)

The inspectors observed conduct of surveillance tests to determine if approved procedures were used, test instrumentation was calibrated, qualified personnel performed the tests, and test acceptance criteria was satisfied. The inspectors verified that the surveillance tests had been properly scheduled and approved by shift supervision prior to performance, control room operators were knowledgeable about testing in progress, and redundant systems or components were available for service, as required. The inspectors routinely verified adequate performance of daily surveillance tests including instrument channel checks, and jet pump and control rod operability tests. The inspectors found the licensee's activities to be generally acceptable.

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The inspector observed the performance of I&C Work Order (WO) R0526192-01, " Unit 3 HCU Rack 'C' Accumulator Instrument PM's."

Two I&C technicians were performing the surveillance locally at the hydraulic control units (HCUs) in the reactor building; a third technician was in the control room. When the inspector asked to review the associated work order, it was determined that the work order was neither at the work site nor in the control J

room, but in the foreman's office. The only document in the control room was the individual calibration data sheets. In addition, The controlling procedure (IC-11-00574, " Calibration Procedure for G. E. HCU Accumulator Level Switches") requires the N, charging connection cap to be torqued to 200 in-lbs, and to record the torque value. The WO directs the technician j

to torque the N charging connection cap to a minimum of 200 in-lbs, and the calibration data i

sheet requires the charging cap to be torqued to 150-200 in-Ibs. Neither the WO nor the data sheet required the torque value to be recorded.

Peach Bottom Technical Specification, Section 6.8.1, requires written procedures to be l

impleniented and that procedures will meet the requirements of Section 5.1 of ANSI

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N18.7-1972, which states: if documentation of an action is required, the procedure should be present and followed step-by-step, and necessary data should be recorded as the task is i

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performed. (Violation 50-277 & 50-278/93-15-02).

6.0 MAINTENANCE ACTIVITY OBSERVATIONS (62703)

The inspectors observed portions of ongoing maintenance work to verify proper implementation of maintenance procedures and controls. The inspectors verified that the licensee adequately

i implemented administrative controls including blocking permits, fire watches, and ignition source and radiological controls. The inspectors reviewed maintenance procedures, action requests

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(AR), work orders (WO), item handling reports, radiation work permits (RWP), material certifications, and receipt inspections. During observation of maintenance work, the inspectors verified appropriate Quality Verification (QV) involvement, plant conditions, TS LCOs, equipment alignment and turnover, post-maintenance testing, and reportability review. The inspectors found the licensee's activities to be acceptable.

6.1 E-33 Bus Outage During the Unit 3 mini-outage, the licensee replaced the 4160/480 VAC step-down transformer for emergency load center E-334. This transformer was the first of eight emergency load center transformers to be replaced under Modification 5099, " Replace Class IE Load Centers." The

modification (MOD) replaces the existing 500kVA, gas-filled ITE transformers with 1000 kVA, j

dry-type Asea Brown Boveri (ABB) transformers. The objective of the MOD is to improve the post-LOCA voltage regulation throughout the Class IE electrical distribution system. The inspector reviewed the MOD package, discussed the MOD with several representatives of the licensee's staff, and observed maintenance activities during the transformer replacement in the field.

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Recent voltage regulation studies identified that unequal loading existed between the IE load centers. This condition has resulted in elevated temperatures inside the transformers which has degraded winding insulation. The licensee is concerned that the transformers have prematurely approached the end of their qualified life. The existing emergency load center transformers have a very small margin for load growth. Administrative controls have been necessary to control the loading on these transformers in order to maintain proper voltage regulation for the 480 VAC safety-related electrical distribution system.

The new design transformers are Class IE, and are environmentally and seismically qualified.

l The increased capacity of the transformer lowers the heat generated. In addition, during LOCA conditions, transformer output voltage regulations is improved during large motor starts.

Transformer life is thereby increased and the need for supplemental cooling is eliminated.

The installation of the new transformer does change the facility's Updated Final Safety Analysis Report (UFSAR). The MOD revises station electrical diagrams described in UFSAR Chapter l

8. The inspector reviewed these changes and found them to be appropriate. The inspector assessed the 10CFR50.59 review and determined that the MOD does not involve an unreviewed

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safety question. The inspector concluded from his review of the MOD package that the MOD would enhance the IE electrical distribution system.

The inspector observed the removal cf the old transformer and installation of the new transformer at load center E-334. All work was performed in accordance with the work order and the licensee's electrical safety procedures. The inspector noted that proper grounding f

devices were in place, that potential transformer fuses were appropriately removed, and blocking clearances were correctly hung. The transformer removal and replacement activities were well planned, supervised, and completed safely.

The transformer installation requires a unit outage. The licensee has established a schedule to replace two transformers per unit refueling outage beginning with the 1993 Unit 3 outage and continuing until the 1996 Unit 2 outage. The licensee plans to replace the second transformer for load center E-134 during the upcoming refueling outage.

The inspector reviewed the modification acceptance test (MAT) for the replacement of the E-334 transformer.

Each transformer installed under MOD 5099 will undergo a MAT.

The satisfactory completion of the MAT ensures operability of the new transformer. The MAT includes core insulation tests, phase energization and relay testing, and a temperature run test.

During the temperature run test, the transformer is gradually loaded while allowing the tempera-

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tures to stabilize. Corresponding high load temperatures are calculated for 100% and 150%

percent loading. The inspector reviewed these temperature calculations and found them to be adequate. The test results for MAT 5099-1 for the E-334 emergency load center transformer, were submitted for the Plant Operation Review Committee's (PORC) review. The PORC

,

declared the load center operable on July 11,1993. The inspector had no further questions.

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6.2 Drywell Tour The inspector toured the Unit 3 drywell following completion of the major outage maintenance j

activities. The tour was intended to review the material condition of the drywell, evaluate

!

housekeeping performance following drywell maintenance activities, and to confirm the licensee's response to NRC Bulledn (BL) 93-02.

NRC Bulletin 93-02, " Debris Plugging of Emergency Core Cooling Suction Strainers," alerted licensees to the potential for loss of net positive suction head for the emergency core cooling systems (ECCS) due to accumulation of fibrous insulating material and debris capture resulting

t in blocking of the suction strainers. The BL requested licensees to identify any fibrous material installed or stored in the containment and to take compensatory measures to assure the functional i

capability of the ECCS.

'

l The licensee's rcyuuse to BL 93-02 stated that no temporarily installed fibrous material was l

utilized in the drywell. Additionally, the licensee stated that plant procedures require an inspection of the drywell following outages when there is access to the drywell. During the

'

tour, the inspector observed that the licensee's response appeared accurate.

While reviewing the condition of the drywell components, the inspector observed that insulation

'

was tightly installed, valve condition was good, and that component preservation was good. The licensee identified and appropriately addressed two minor valve discrepancies. The inspector determined that the overall condition of the drywell components was good.

l The inspector observed the licensee's performance of the drywell inspection and noted that the j

j pre-job radiological briefing was good, personnel and material access control to the drywell was appropriate, and that the licensee personnel who performed the inspection were knowledgeable.

The licensee performed a thorough inspection and identified and promptly corrected a few minor i

,

!

housekeeping deficiencies. The inspector concluded that drywell housekeeping and the licensee's inspection were acceptable.

i 6.3 Operability of 480 VAC Circuit Breakers On July 27, 1993, during an inspection of 480 volt A/C breakers for training purposes, an auxiliary operator (AO) noticed that the terminal block hold-down screws were missing. Further inspection of other 4S0 volt A/C breakers and similar 250 volt D/C breakers identified several other instances of missing hold-down screws. The licensee was able to replace the missing screws on all but one breaker. The breakers were not in an analyzed condition with respect to seismic qualification. Until such time as the screws were replaced, the associated safety systems were declared inoperable.

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The licensee took immediate steps to minimize the effects of the missing screws, and initiated an investigation (RE/EIF '2-93-313) to determine the extent of the problem, the cause, and the long-term corrective actions. The inspectors have no additional questions until the investigation is complete. This will remain an unresolved item (URI 50-277 & 50-278/93-15-03).

6.4 Limited Senior Reactor Operators During preparations for the Unit 3 mini-outage, it was discovered that the senior reactor operators limited to fuel handling (LSROs) were reactivating their licenses by standing an under-instruction watch in the control room. 10CFR55, " Operators' Licenses," subpart 55.53(e),

states that an operator must stand a minimum of seven 8-hour or five 12-hour shifts per quarter to maintain an active license. If an operator does not maintain an active license, subpart 55.53(f) details the necessary steps to reactivate a license: an LSRO must stand one shift under the direction of an active SRO/LSRO and in the position to which the LSRO will be assigned.

The LSROs are only licensed to stand watch on the refuel floor; thus, standing an under-instruction watch in the control room vice on the fuel floor was not adequate to reactivate their licenses. The LSROs are dual licensed at both Peach Bottom and Limerick. The inspector discussed this anomaly with the licensee. This will remain an unresolved issue pending further review by the NRC. (URI 50-277 & 50-278/93-15-04)

7.0 RADIOLOGICAL CONTROLS (71707)

The inspectors examined work in progress to verify proper implementation of health physics (HP) procedures and controls. The inspectors monitored the ALARA (as low as reasonably achievable) program implementation, dosimetry and badging, protective clothing use, radiation surveys, radiation protection instrument use, handling of potentially contaminated equipment and materials, and compliance with RWP (radiation work permits) requirements. The inspectors observed that personnel working in the radiologically controlled areas were meeting applicable requirements and were frisking in accordance with HP procedures. During routine tours of the units, the inspectors verified a sampling of high radiation doors to be locked, as required. In general, activities monitored by the inspectors were found to be acceptable. However, during the Unit 3 outage, the inspectors observed several instances that were not consistent with HP procedures:

Inside the RWP area, two individuals were placing yellow plastic over the rail that

surrounds the reactor cavity. It was noted that neither individual had the anti-contamination protective clothing closed in the chest / neck area. This is not consistent with procedure i

HP-C-510, " Selection, Use, and Control of Protective Clothing," section 7.2.1.

.

The RWP area on the refuel floor was designated by yellow-magenta tape on the floor and

yellow-magenta rope supported by stanchions about 3 foot off the floor. The appropriate radiological warning signs designating a potentially contaminated area were hung from the rope. The tape was about 12 inches inside of the rope. Procedure HP-215, " Establishing j

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and Posting Controlled Areas," section 7.1.1, requires postings of all boundaries to be consistent. A board lay on the floor inside of the rope barrier but outside of the tape

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barrier. A worker, outside of the RWP area, picked up the board to move it to another l

area. When questioned by the NRC inspector as to the radiological condition of the board, the worker stated that he thought the tape was the " official" barrier and the rope was a

'

warning.

Personnel exiting the RWP area on the refuel floor were not consistent in removing their

anti-contamination clothing. Specifically, both sets of gloves were removed before removal of the plastic shoe covers, however, HP-C-510, Exhibit 2, steps 9 and 10, require the plastic booties to be removed prior to removal of the plastic gloves.

Because the spent fuel pool and the reactor cavity are within the RWP am.a. accountability

procedures require personnel and equipment to be logged into the area. A resiew of the accountability log identified that an HP technician had failed to sign out of the log after exiting the area. This is not consistent with procedure A-C-130, " Fuel Floor Material and Personnel Accountability," section 7.8.1, which states that personnel accountability shall be maintained.

The NRC inspector identified the discrepancies to the HP and maintenance foremen in the area;

appropriate and immediate corrective actions were taken for the specific deficiency identified.

Discussions with HP and plant management focused on the number of observations, the lack of

'

supervision in the field, and the need for additional attention to detail in the area of radiological work practices. These deviations from the procedures are a violation of Peach Bottom Technical Specification 6.8.1, in that procedures shall be implemented, as written. (Violation 50-277 &

50-278/93-15-05).

8.0 PHYSICAL SECURITY (71707)

,

The inspectors monitored security activities for compliance with the accepted Security Plan and associated implementing procedures. The inspectors observed security staffing, operation of the Central and Secondary Access Systems, and licensee checks of vehicles, detection and assessment aids, and vital area access to verify proper control. On each shift, the inspectors observed protected area access control and badging procedures. In addition, the inspectors l

routinely inspected protected and vital area barriers, compensatory measures, and escort procedures. The inspectors found the licensee's activities to be acceptable. The NRC issued an information advisory to licensed facilities recommending that security awareness be heightened. The inspectors reviewed the licensees's response to this advisory and concluded that their response was appropriat....

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9.0 PREVIOUS INSPECTION ITEM UPDATE (92701, 92702)

I (Open) URI 92-07-02. "2B Condensine Chamber Desien Problem" l

During March 1992, due to problems with the reactor vessel level instrumentation served by the 2B condensing chamber and reference leg, the licensee declared several safety systems inoperable and shutdown Unit 2. The licensee installed a temporary modification to assess condensing chamber performance and developed acceptance criteria for evaluating divergence between instrument channels. The NRC concluded that the licensee was adequately monitoring the affected instrumentation and pursuing a long term solution. This item was left unresolved pending assessment of the licensee's actions to correct this problem.

l l

The inspector reviewed the licensee's plans for installing a permanent hardware modification to ensure reliable reactor vessel water level indication and concluded that the licensee's plans for resoking this issue are acceptable.

This item remains open pending the licensee's

,

demonstrations during plant operation that this modification has resolved the original level l

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divergence problem.

I (Closed) URI 92-80-01. " Assessment of Inonerable Control Room Instrumentation" During the Integrated Performance Assessment Team (IPAT) inspection, the Team identified three instances in which the effect of inoperable control room instrumentation had not been effectively evaluated with respect to emergency operating procedure implementation. The Team expressed concern for the total number of inoperable instruments, the cumulative effect of the inoperable equipment on operator and plant response to transients, and the effectiveness of evaluations of inoperable instrumentation.

The inspector reviewed the licensee's actions in response to these issues. The licensee revised the equipment control section of the Operator's Manual to formalize guidance regarding identification and tracking of equipment problems that present a significant operational concern.

The inspector interviewed selected personnel who are required to implement this guidance and determined they were knowledgeable of the requirements. The inspector noted, during control panel walkdowns, that implementation of this guidance was generally good; however, some minor inconsistencies existed regarding the significance assigned to different equipment problems. The inspector discussed this issue with the Operations Support Supenrisor who indicated that this issue had already been identified by them and they were in the process of reviewing and enhancing this guidance.

The inspector reviewed the backlog of outstanding control room instrument deficiencies and noted that the licensee has made progress in reducing the number of outstanding deficiencies.

The licensee actively tracks inoperable control room instrumentation status and has assigned an individual from the maintenance planning department to coordinate resolution of these

,

deficiencies. The inspector concluded that the licensee's actions to resolve this issue were l

l appropriate. Unresolved item 92-80-01 is closed.

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l (Closed) URI 92-80-02. " Interim Self-Assessment Corrective Actions"

'

During the IPAT inspection, the Team reviewed the findings of a licensee self-assessment. The Team noted that many of the findings required extensive improvement programs; however, several of the findings warranted immediate corrective actions to reduce the potential for future safety problems.

The inspector reviewed procedure LR-C-10, " Investigation ofIn-House Events." This procedure formalizes guidance regarding investigation of self-identified weaknesses. The inspector noted l

l that the procedure was enhanced in January 1993, requiring interim corrective actions be l

identified to prevent recurrence of an event, while the final corrective actions are being developed. The inspector reviewed selected licensee event reports and noted that appropriate interim corrective actions were identified. The inspector was satisfied with the licensee's response to this issue. This unresolved item is closed.

(Closed) URI 92-80-03. " Assessment of installed Instruments Found Out of Commission" During the IPAT, the Team noted that the licensee lacked pmcedures to ensure that permanently installed instrumentation found to be out of calibration is properly assessed for the effect on related system operability.

The licensee developed a program to establish action guidelines to be followed upon discovery that an installed plant instrument is out of tolerance. The inspector reviewed the program requirements specified in AG-93, "Out of Tolerance Notification and Disposition of Installed

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Plant Instrumentation," and did not identify any concerns with the program. The inspector also interviewed the program coordinator, reviewed program documentation, and concluded that the licensee was properly implementing the program. This unresolved item is closed.

10.0 MANAGEMENT MEETINGS (71707,30702)

10.1 Performance Review A meeting between licensee and NRC management was held on July 21, 1993, at the Site Management Building in Delta, Pennsylvania. The meeting was held to discuss Peach Bottom's self-assessmeat of their performance for the first half of the current Systematic Assessment of Licensee Performance (SALP) report period. The presentation by the licensee and the dialogue

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between NRC and licensee management was constructive and informative. The information l

enabled the NRC to better understand management activities and initiatives at Peach Bottom.

l The licensee's presentation material is included as Enclosure 1.

I l

On July 6, the inspectors met with consultants for the licensee's Nuclear Committee of the Board of Directors. The meeting was held to exchange information regarding Station Performance.

The meeting enabled the inspectors to better understand the licensee's off-site assessment capabilities.

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10.2 Exit Meeting The Resident Inspectors provided a verbal summary of preliminary findings to the station management at the conclusion of the inspection. During the inspection, the Resident Inspectors verbally notified licensee management concerning preliminary findings. The inspectors did not provide any written inspection material to the licensee during the inspection. The licensee did not express any disagreement with the inspection findings. This report does not contain proprietary information. The following specialist inspections also occurred during the report period:

Date Subiect Report No.

Inspector 6/17-23 Radwaste and Transportation IR 93-13 Ecken 6/21-25 Maintenance Troubleshooting IR 93-14 Bower i

j 6/28-30 Emergency Preparedness Exercise IR 93-10 Laughlin 7/12-16 Erosion /Corrosien IR 93-16 McBrearty

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HEALTH PHYSICS

'

i

improved Radworker Practices e

!

,

Aggressive Exposure Reduction i

'

!

Significant Reduction in Personnel Contaminations

i j

Health Physics Teams Visit Other Utilities

~

l Increased Health Physics Staff Technical Expertise

,

k

?

!

.

-

.

-.

-

.'.:

.

-

.

'

CHEMISTRY Reduction in Liquid Releases e

Maintain Water Quality Early Identification of Fuel Failure

-

i l

Previous SALP Data 1991-1992 " Performance in the Areas of l

-

l Effluent Controls and the REMP Continued to be Excellent"-

-

~

4

.

-. -

-

.

. -

- -

-

.

-

.

- -

-

- - - - - -

-

.

-

.

-

.

.;

.

-

.

RADWASTE

'

< Station Decontamination e Radwaste Volume Reduction

Continued Excellence in Shipping and Transportation

.

o Previous SALP Data

,

-

1991-1992 "Overall Performance... Remained Excellent"

>

j Leak Reduction

.

Drip Bag Controls

!

' Painting

-

.

l e Area Decon Where Ops Frequents

.

-

- -

- - -

..

-

-.

..

.

. -...

.

...

.

..

.

..

.

_

.

I IMPROVEMENT INITIATIVES

e Advanced Rad Worker Program

o Drywell Shielding

!

e Dose Reduction Plans

.

l l

e Robotics

!

. Cond. Demin. Performance

.

!

e Irradiated Hardware Disposal i

j e Reduction of Outside Radioactive Material Storage

l i

- -

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CONTINUOUS IMPROVEMENTS e Non-Outage CM Backlog i

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.

Control Room Equipment

>

!

2a ESF Actuations

.

i

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.-

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'

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PBAPS ACTION REQUESTS Non-Outage CM Backlog

"

~~,,___s

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.,,_ _ _ _ _ _ _ _ _ _ y ~ ~

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-

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-

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Total 6 Goal Note: 57% over 90 days old

.-

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CM BACKLOG OVER 1 YEAR OLD

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CONTROL ROOM EQUIPMENT Blue Dots, Blue Tags, Beige Tags 150

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IMPROVEMENT INITIATIVES

,

e Live Time Planning

e Backlog Reduction J

-

NEEDS

-

Focus on " Minor Work"

.

!

l Maintenance Technician Program

i

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,

WORK MANAGEMENT Material Condition of Plant

.

Continuous Improvements

-

Plant Painting & Decontamination

!

-

Less Plant Transients Caused by Equipment Malfunctions

-

Reactor Chemistry Maintained at Higher Standards

.

-

Condenser Replacement llBoth Units's

,

-

Feedwater Heater Replacements - U/2

.

-

Full Core Display - Digital Readouts

.

,

1

- - -

- -.

-.. -.

..

--

-.

.

..

,

.

_

.

WORK MANAGEMENT

,

Material Condition of Plant

,

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l

-

Feedwater Control Valve Positioner and Linkage

.

Replacement

.

-

Moisture Separator Dump and Drain Valve

,

Replacement

-

Main. Steam Isolation Valve Poppet Upgrade i

-

New Radwaste Pumps

!

-

Circulation Water Pump Overhaul Scheme ~

,

!

\\

-

- -

- - -

- -

..

- - -

-

.

.

.

-

,

.

,

WORK MANAGEMENT Material Condition of Plant

,

!

a Continuous Improvements [ Cont.)

-

EHC System Improvements

-

Chem Lab Modification

-

Digital Feedwater Control Modification

.

l

-

Recirculation Pump Control System Modification

-

Site Building Improvements i

j i

!

.

. -

-

- - - - -

.

- -

-

- - -

.

-

.

_

.

.L

'

WORK MANAGEMENT Capital Investment a1991 78.4 Million ll Actual?

.

a 1992 63.8 Million

' Actual}

?1993 73 Million

l a1994 72.4 Million

< 100% Direct Cost Only? Excludes AFUDC & O.verhead.

'

Includes Above the Line & Below the Line Projects.

!

-

.

-..

.

.. -, -..

_

-

.

,

-

.

WORK MANAGEMENT

!

Improvement Initiatives

,

Auxiliary Boiler Replacement O Complete?

? Large Case Recorder Replacements in Control Room Cooling Tower Work

-

'

Water Plant Modification

.

,

a Reactor Water Clean Up Modification l

Containment Atmospheric Dilution Modification n

!

l Plant Heating Modification t

Offsite Power Source Modification I

_

.

-

__

_-

.

..

.,

.

'

.

'

WORK MANAGEMENT Station Modification : Management Group (S M MG?

.

Members:

G. Cranston G. Edwards G. Gellrich

,

D. LeQuia J. McElwain l

D. Meyers SMMG Que after safety significant issues weighted towards

,

operational impact issues, balance of plant included Added plant reliability and reducing cost in bot, dose and dollars

]

to Operations and Maintenance.

- -

-

.

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.

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a WORK PLANNING l SCHEDULING

'

f Work packages produced by Maintenance Planning

-

and Contractor of Choice for Daily and Outage Work Scheduling provided for daily activities including

-

testing, corrective maintenance, preventive maintenance and non-outage modification work based on 12-week rolling schedule l

Scheduling for Refuel and Forced Outages is

-

,

Performed Using the Same Scheduling Tools

.

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' System Improvement Initiatives

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Performance Monitoring

a System Manager Role i

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CONTINUOUS IMPROVEMENTS

Community Compact e Tour Program

,

Partnership Program

a Educational Outreach Program

.

.-

..

-.

.;

..

.

-

.

i CONTINUOUS IMPROVEMENTS

'

VP's Quarterly Communications to

-

Community Leaders

.

Annual Meetings with Government Leaders Community Liaison Program l

~

Speakers Bureau Program

.

!

i

,

.

.

PERFORMANCE LVDICATOR 1993 Educational Outreach Program

.

$ - 3'

-

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Jan Feb Mar Apr May Jun Jul Aug Sep Oct t4ov Dec

of Students Reached Goal (1,020)

-

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Program involves Career Days, Class Lectures and School Assemblies About the Nuclear Power Industry

...

..

.

..

.

.;

.

?

PERFORMANCE INDICATOR

Tour Program

'

,

Tours 1992 Tours 1993

,

300 300

,

250

-

250

-

[Q"7 blg2 f

bb lgff 200

-

200

-

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,

.

,

.

'

,

.

~

WHAT OUR AUDIENCES ARE SAYING ABOUT US

... On the VP's Letters to Community Leaders

"It is good to read that Peach Bottom is turning things around so nicely. Aberdeen Proving Ground is in my district so I know how important it is to work with the community in an effort to solve problems. I am glad that the Philadelphia Electric Company is making this effort."

U.S. Representative

Helen Delich Bentley

,

-

--

- -

.

.

,

.

k i

... ON THE TOUR PROGRAM

.

i

"Your presentation to our group and the tour of PBAPS were both interesting and educational. Feedback from the Board members determined that all those involved gained a great deal of information and answers to many of their questions regarding PBAPS and emergency response operations. The tour provecl to be quite beneficial and enjoyable for all of the participating

members."

i l

'

Robert H. Hiscock, Acting Chairman Radiation Control Advisory Board

'

State of Maryland

'

Department of the Environment

-

.

'

,

k

.

. -. - -

- - -

. -

-

. - - -.

- -

-

. -. - -

-

.

'

-

,

.

. ON THE COMMUNITY COMPACT

~

..

I applaud your efforts in this endeavor and

"

wish you every success with it."

Delegate Mary Louise Preis State of Maryland

.

-

.

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.

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IMPROVEMENT INITIATIVES

-

Expand the Speakers Bureau Program

-

'

- Increase the Quality and Quantity of

.

Educational Outreach Programs

,

Survey Target Audiences to Determine What i

They Want from Peach Bottom (Late 1993?

,

a

'

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.

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PEACH BOTTOM IS IN A MODE OF CONTINUOUS IMPROVEMENT

.

- We're doing well a Other utilities from the U.S. and the world

!

come here to see how we do business Still have areas to improve l

)

e We will never be done, we'll always be

continuously improving i

i

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- _. _ _ -. _ - _ _.... _ _. _ -.

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,

o

,

.

WHAT DOES THE FUTURE LOOK LIKE

.

AT PEACH BOTTOM?

- A good place to work

-

High morale

~ People practicing behaviors that allow us

to continually improve

-

k

'

. -.

. - -.

. -..

..... -.

- -

-............

v t e Inspection Report - Peach Bottom - 1993015
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1993 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 7(0) 8(0) 9(0) 11(0) 12(0) 13(0) 14(0) 15(0) 16(0) 18(0) 19(0) 20(0) 23(0) 24(0) 26(0) 27(0) 28(0) 29(0) 30(0) 80(0) Security Operator Exams Emergency Planning Other

IR 05000277/1993015

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REGION I

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